ML20080K967

Safety Evaluation Supporting Amend 120 to License NPF-29

ML20080K967
Person / Time
Site:	Grand Gulf
Issue date:	02/21/1995
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20080K966	List: ML20080K967
References
NUDOCS 9503020026
Download: ML20080K967 (173)

Text

-

a na k

UNITED STATES :

.p i.

j NUCLEAR REGULATORY COMMISSION

[I

\\, * * * * * /

2 WASHINGTON, D.C. enmaa mam I

i

-SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION i

RELATED TO ANENONENT N0.120 TO i

FACILITY OPERATING LICENSE No. NPF-29 ENTERGY~ OPERATIONS, INC.

l GRAND GULF NUCLEAR POWER STATION DOCKET NO. 50-416-j I.

INTRODUCTION i

The Grand Gulf Nuclear Power Station-(GGNS) currently operates with technical specifications (TS) issued on November 1,1984 with the original operating i

license, Operating License No. NPF-29, as amended from time to time over the years. By letter dated October 15,-1993, and supplemented by letters dated-1 April 26, 1994, November 10, 1994, February 10, 1995 and February 14,-1995, the Entergy Operations, Inc. (the licensee) proposed to amend Appendix A of -

the license to revise, in its entirety,' the GGNS technical ~ specifications.

4.

The proposed amendment was based on NUREG-1434, " Standard Technical.

i Specifications - General Electric Plants, BWR/6," issued in September 1992, I

and on guidance provided in the Commission's " Final Policy Statement on i

Technical Specifications Improvements for Nuclear Power Reactors" (" Final' i

i Policy Statement"), published on July 22,1993 (58 FR 39132). The overall i

[

objective of the proposed amendment, consistent with the NRC's Final Policy i

Statement, was to completely rewrite, reformat, and streamline the existing i

I GGNS technical specifications.

3 1

i In addition to basing its improved TS on NUREG-1434 and the Commission's Final

{

Policy Statement, the licensee used portions of the existing TS as a basis for the GGNS improved TS. Plant-specific issues,' including plant-unique design features, plant-unique requirements, and plant-unique operating practices were j

[

discussed with the licensee during a series of meetings concluding on August

}.

18, 1994.

In addition,-meetings were held with the Owners Groups to discuss' matters of a generic nature that were not incorporated in NUREG-1434; these-i generic issues were considered for specific applications in the GGNS improved i

TS. Consistent with the Commission's policy statement, GGNS proposed 2

transferring some TS requirements to other licensee-controlled documents.

In i

addition, emphasis was'placed on human factors principles to add clarity and l

understanding to the GGNS improved TS and to define more clearly the i

appropriate scope of the TS.

Further, significant changes were proposed to j

the Bases section of the GGNS TS to enhance the clarity and understanding of j

each specification.

i I

t In addition to the original.0ctober 15, 1993 submittal, the licensee has I

j submitted, and the staff has accepted, a number of changes to the existing-i GGNS TS. The review and approval of these TS amendments was independent of the GGNS improved TS review effort. These previous TS changes are reflected, l

as appropriate, in the GGNS improved TS. This Safety Evaluation (SE)-

describes only those TS changes which affected implementing the GGNS improved i

TS.

i 9503020026 950221

~

l PDR ADOCK 0 4y6 P

i d,,

1 4

2 9

The Commission's proposed action on the GGNS license amendment request was published in the Enderal Reaister on April 12, 1994 (59 FR 17404) and April 21, 1994 (59 FR 19031). Changes in the licensee's proposed TS that resulted

)

i from discussions with the licensee during the staff's review are discussed in this SE. These plant-specific changes serve to clarify the TS with respect to the guidance in the Commission's policy statement and NUREG-1434. Therefore, the changes are within the scope of the action described in the initial.

j Federal Reaister notice.

l During its review of the GGNS license amendment application, the NRC staff relied on the NRC's Final Policy Statement and on NUREG-1434. This SE documents the basis for the staff's conclusion that GGNS can convert its j

existing TS to those based on NUREG-1434, as modified by plant-specific changes, and that the use of the GGNS improved TS is acceptable for continued plant operation. The staff also acknowledges that, in accordance with the l

Commission's policy. statement, the conversion to the STS is a voluntary i

process. Therefore, the GGNS improved TS reflect some differences that i

correspond to the existing licensing basis for the plant. The staff has identified the changes to the existing GGNS TS and has included an explanation i

of the significant changes in this SE.

Individual section topics and the corresponding section numbers are identical to those given in NUREG-1434.

i j

For the reasons stated infra in this SE, the staff finds that the TS issued l

with this license amendment satisfy Section 182a of the Atomic Energy Act, 10 CFR 50.36, and the guidance in the Commission's Final Policy Statement, and 1

that they are in accord with the common defense and security.and provide adequate protection to the public health and safety.

5 a

i II.

BACKGROUND Section 182a of the Atomic Energy Act requires that applicants for nuclear i

power plant operating licenses shall state:

[

(S]uch technical specifications, including information of the 1

amount, kind, and source of special nuclear material required, the place of the use, the specific characteristics of the facility, i

and such other information as the Commission may, by rule or regulation, deem necessary in order to enable it to find that the

{

utilization... of special nuclear material will be in accord with the common defense and security and will provide adequate 1

protection to the health and safety of the public. Such technical l

specifications shall be a part of any license issued.

l In 10 CFR 50.36, the Commission established its regulatory requirements related to the content of TS.

In doing so, the Commission place,d emphasis on i

i those matters related to the prevention of accidents and those matters related j

to the mitigation of accident consequences; the Commission roted that applicants were expected to incorporate into their TS "tho',e items that are i

directly related to maintaining the integrity of the physical barriers l

designed to contain radioactivity." Statement of Consideration, " Technical Specifications for Facility Licenses; Safety Analysis Reports," 33 Fed. Reg.

e o

1 l'

y 3

18610 (December 17,1968).

Pursuant to 10 CFR 50.36, TS are required to j

include items in five specific categories, including (1) safety limits, i

limiting safety system settings and limiting control settings; (2) limiting 3

i conditions for operation; (3) surveillance requirements; (4) design features;

~1 and (5) administrative controls. However, the rule does not specify the i

particular requirements to be included in a plant's TS.

For several years, the NRC and industry representatives have sought to develop 9uidelines for improving the content and quality of nuclear power plant TS.

On February 6, 1987, the Commission issued an interim policy statement on TS i

improvements, " Proposed Policy Statement on Technical Specification i

Improvements for Nuclear Power Reactors" (52 FR 3288). During 1989 through 1992, the utility Owners Groups and the NRC staff developed improved Standard Technical Specifications (STS) that would establish models of the Commission's i

policy for each primary reactor type.

In addition, the staff, licensees, and j

the Owners Groups developed generic administrative and editorial guidelines in i

the form of a " Writers Guide" for technical specifications, which affords a significant enhancement of human' factors considerations and was used throughout the development of licensee-specific improved TS.

In September 1992, the Commission issued NUREG-1434, which was developed i

utilizing the guidance and criteria contained in the Commission's interim i

policy statement.

It was established as a model for developing improved i

technical specifications for the BWR/6 plants in general and for the improved

}

Grand Gulf Nuclear Station TS specifically. NUREG-1434 reflects the reruits of a detailed review of the application of the interim policy statement criteria to generic system functions, which were published in a " Split Report" i

issued to the NSSS Owners Groups in May 1988.

NUREG-1434 also reflects the results of extensive discussions on various drafts of standard technical 4

specifications, so that the application of the TS criteria and the Writers Guide would consistently reflect detailed system configurations and operating characteristics for all NSSS designs. As such, the generic Bases presented in NUREG-1434 provide an abundance of information regarding the extent to which the standard technical specifications present requirements which are necessary i

to protect the public health and safety.

J j

On July 22, 1993, the Commission issued its Final Policy Statement. Therein, the Commission e.tpressed its view that satisfying the guidance in the policy 4

statement also satisfies section 182a of the Atomic Energy Act and 10 CFR 3

i 50.36. The Final Policy Statement described the safety benefits of the l

improved STS and encouraged licensees to use the improved STS as the basis for i

plant specific TS amendments, and for complete conversions to improved STS.

j Further, the Final Policy Statement provided guidance to evaluate the required 4

scope of the technical specifications, and finalized the guidance criteria to be used in determining which of the design conditions and associated i

j surveillances need to be located in the TS. The Commission noted (58 FR i

at 39136) that, in allowing certain items to be relocated to licensee-controlled documents while requiring that other items be retained in the TS,

)

it was adopting the qualitative standard enunciated by the Atomic Safety and j

Licensing Appeal Board in Portland General D ectric Co. (Trojan Nuclear i

Plant), ALAB-531, 9 NRC 263, 273 (1979). There, the Appeal Board observed:

1

e o

4

[T]here is neither'a statutory nor a regulatory requirement that every operational detail set forth in an applicant's safety

.l j

analysis report (or equivalent) be subject to a technical j-specification, to be included in the license as an absolute condition of operation which is legally binding upon the licensee unless and until changed with specific Commission approval.

Rather, as best we can discern it, the contemplation of both the Act and the regulations is that technical specifications are to be reserved for those matters as to which the imposition of rigid conditions or limitations upon reactor operation is deemed l

necessary to obviate the possibility of an abnormal situation or i

i event giving rise to an immediate threat to the public health and J

safety.

i In accordance with this approach, existing TS requirements which fr.11 within or satisfy any of the criteria in the Final Policy Statement should be retained in the TS, while those TS requirements which do not fall within or l

4 satisfy these criteria may be relocated to other licensee-controlled J

documents. The Final Policy Statement criteria are as follows:

i l.

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

l 2.

A process variable, design feature, or operating restriction that is an 1

initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity j

of a fission product barrier.

3.

A structure, system, or component that is part of the primary success i

path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge J

to the integrity of a fission product barrier.

i i

4.

A structure, system, or component which operating experience or probabilistic safe}y assessment has shown to be significant to public j

health and safety.

1 j

In its license amendment application, the licensee proposed changes to i

existing TS requirements using the Final Policy Statement and NUREG-1434 as guidance. Changes to NUREG-1434 were also proposed by the licensee due to

}

The Commission recently promulgated a proposed change to 10 CFR 50.36, pursuant to which the rule would be amended to codify and incorporate these i

criteria (59 FR 48180). The Commission's Final Policy Statement specified that Reactor Core Isolation Cooling, Isolation Condenser, Residual-Heat Removal, Standby Liquid Control, and Recirculation Pump Trip are to be included in the TS under Criterion 4.

In the proposed change to 650.36, the t

Commission specifically requested public comments regarding application of i

Criterion 4.

Until additional guidance has been developed, Criterion 4 is not j

being applied to add TS restrictions other than those indicated above.

d

)

e e.

l 5

differences between the plant-specific licensing basis and the design basis provided in Bases in NUREG-1434.

In this SE,-the licensee's proposed changes to its existing TS requirements are grouped into four general categories as follows:

administrative, i.e.,

non-technical changes; relocated requirements, i.e., movement of requirements from existing TS (an NRC-controlled document) to specified licensee-controlled documents; more restrictive requirements, i.e., additions to existing TS; and i

less restrictive requirements, i.e., relaxations to, or deletions from existing TS requirements. These four general categories of changes to the licensee's existing TS requirements may be better understood as follows.

Administrative Chanaes Non-technical, administrative changes were intended to incorporate human-factors principles into the form and structure of the improved plant 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 of requirements without affecting technical content or operational l

requirements. Every section of the proposed TS reflects this type of change.

In order to ensure consistency, the NRC staff and the various licensees of the j

BWR/6 conversion plants have used NUREG-1434 as guidance to reformat and make 1

other administrative changes. The licensees proposed such changes as-(a) providing the appropriate numbers, etc., for NUREG-1434 bracketed information (information which must be supplied on a plant-specific basis, and

)

which may change from plant to plant), (b) identifying plant-specific wording for system names, etc., and (c) changing NUREG-1434 section wording to conform to existing licensee practices.

The staff has reviewed all of the administrative and editorial changes proposed by the licensee (or generically by the licensees) and finds them acceptable, since they are compatible with the " Writers Guide" and NUREG-1434, and are consistent with the Commission's regulations. The-non-technical administrative changes are discussed individually in this evaluation.

Relocated Reauirements As summarized above, the Commission's policy statement provides that existing i

TS requirements which do not satisfy or fall within any of the four specified criteria may be relocated to appropriate licensee-controlled documents.

In the licensee's application, such requirements are generally relocated to the i

Updated Final Safety Analysis Report (UFSAR) and TS Bases. Unless otherwise specified in this safety evaluation, the relocated limiting conditions for operation (LCO) portion t

• the existing TS, which includes the system description, design limits, functional capabilities, and performance levels, will be relocated to the UFSAR. The relocated provisions of the existing TS action statements and surveillance requirements will be relocated to the UFSAR or TS Bases depending on the nature of the requirements being relocated. Any time the operability of'a system or component has been affected by repair, maintenance or replacement of a component, plant procedures require that a post maintenance test be performed to demonstrate operability of the system or

1 1

l

^

l 1

6 i

component. The existing TS have various post-maintenance surveillance requirements distributed throughout which have been relocated from the improved TS.

In addition the details and methods of operation of a system j

during the performance of a surveillance have been relocated from the existing TS.

Examples include descriptions of tests to assure controls of the system are operable, controls during functional testing of components, and setpoint verification which inherently performs a functional test of the instruments and the cycling of valves. These procedures will similarly be described in the UFSAR or TS Bases. The requirements which are being. relocated from the existing plant TS to licensee-controlled documents are summarized in Table 1.

a The facility and procedures described in the UFSAR and TS Bases can only be 1

revised in accordance with the provisions of 10 CFR 50.59, which ensures an i

auditable and appropriate control over the relocated requirements and any 1

]

future changes to these provisions. Other licensee-controlled documents 3

include provisions for making changes consistent with other applicable j

regulatory requiremeni.s; for example, the Offsite Dose Calculation Manual (00CM) can be changed in accordance with 10 CFR Part 20; the Emergency Plan Implementing Procedures (EPIP) can be changed in accordance with 10 CFR 1

50.54(q); and the administrative instructions' that implement the Quality j

Assurance Manual (QAM) can be changed in accordance with 10 CFR 50.54(a) and 10 CFR Part 50, Appendix B.

Temporary procedure changes are also controlled by 10 CFR 50.54(a).

l-l Although the UFSAR already includes most of the design information described j

above, by letter dated November 10, 1994 the licensee committed to confirm 1

that these details are appropriately reflected in the UFSAR, improved TS Bases 4

i or will be included in the next update of these documents. The licensee has also committed to maintain an auditable record of and an implementation schedule for the procedure changes associated with the development of the i

improved plant-specific TS. The documentation of these changes will be i

maintained by the licensee in accordance with the record retention requirements specified in the QA Plan.

i As described in more detail in this evaluation, the staff concludes that appropriate controls have been identified for all of the requirements that are i

i being relocated from the licensee's TS to licensee-controlled documents.

i Until incorporated in the UFSAR and procedures, changes to the provisions i

being relocated from the TS will be controlled in accordance with the applicable existing procedures that control these documents. The NRC will j

conduct an audit of the relocated requirements following implementation to assure that an appropriate level of control has been achieved. The staff concludes that, in accordance with the Commission's policy statement, i

sufficient regulatory controls exist under the regulations, particularly 10 i

CFR 50.59. Accordingly, the staff concludes that these requirements, as l

described in detail in this evaluation, may be relocated from the TS to the UFSAR or to other licensee-controlled documents as specified herein.

i

j e

l 7

i L

l Nore Restrictive Reauirements a

The licensee's proposed improved TS include certain more restrictive requirements than are contained in the existing TS, which are either more conservative than corresponding requirements in the existing TS, or are additional restrictions which are contained in NUREG-1434 but are not i

contained in the existing TS.

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 i

equipment; and more restrictive surveillance requirements. The more restrictive requirements are discussed individually in Section III of this evaluation.

i Less Restrictive Reautrements J

1 The more significant less restrictive requirements are justified on a i

case-by-case Msis as discussed in Section III of this evaluation. When i

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

In most cases, relaxations previously granted to indivMual plants on a plant-specific basis were the result of (a) i generic NRC acticns, (b) new NRC staff positions that have evolved from technological advancements and operating experience, or (3) resolution of the l

Owners Groups' comments on the improved STS. Generic relaxations contained in i

NUREG-1434 were reviewed by the staff and found to be acceptable because they are consistent with current licensing practices and NRC regulations.

The licensee's 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-1434 and thus provides a basis for these revised TS.

t The following sections explain the staff's reasons for concluding that the i

conversion of the licensee's existing TS to improved TS based on NUREG-1434,

?

as modified by plant specific changes, is consistent with the current plant

}

specific licensing basis, applicable regulatory requirements and guidance of the policy statement, and is acceptable.

III.

EVALUATION j

1.0 Use and Application j

1.1 Definitions The definitions appearing in Section 1 of the GGNS improved TS have been I

reorganized from the existing GGNS TS by deleting the identification numbers associated with each definition and listing them in alphabetical order.

l The following definitions have been retained in the GGNS improved TS.

Some editorial changes have been made so that these defined terms are consistent

{.

with NUREG-1434 and with GGNS plant-specific terminology. The modifications have.been accepted by the licensee and, the resulting definitions do not change the intent of the definitions as found in NUREG-1434. Therefore, we find these definitions acceptable for GGNS.

k

,e

.m=

.y*

> wr.

.w--

I

8 ACTIONS AVERAGE PLANAR LINEAR HEAT GENERATION RATE CHANNEL CALIBRATION CHANNEL CHECK CHANNEL FUNCTIONAL TEST CORE ALTERATION CORE OPERATING LIMIT REPORT DOSE EQUIVALENT I-131 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME END OF CYCLE RECIRCULATION PUMP TRIP (EOC-RPT) SYSTEM RESPONSE TIME ISOLATION SYSTEM RESPONSE TIME LEAKAGE (formerly, IDENTIFIED LEAKAGE, PRESSURE BOUNDARY LEAKAGE and UNIDENTIFIED LEAKAGE)

LINEAR HEAT GENERATION RATE LOGIC SYSTEM FUNCTIONAL TEST MINIMUM CRITICAL POWER RATIO OPERABLE-0PERABILITY RATED THERMAL POWER.

REACTOR PROTECTION SYSTEM (RPS) RESPONSE TIME SHUTDOWN MARGIN (SDM)

STAGGERED TEST BASIS THERMAL POWER TABLE 1.2 New definitions MODE, and L were added to the GGNS improved TS. These new definitions are compatible w,ith changes made throughout the GGNS improved TS to clarify the related requirements and to reduce the likelihood of misinterpretation of the improved TS.

The new GGNS definitions (except L -

Plant-specific wording differences have.)been were also defined in NUREG-1434.

reviewed and do not change the meaning of these definitions.

The licensee, in electing to implement the NUREG-1434 Section 1.0 specifications, proposed a number of less restrictive conditions than are allowed by the existing TS. The more significant conditions are the following:

1.

The phrase "or actual," in reference to the injected signal, has been added to the definition of Channel Functional Test. Some Channel Functional Tests are performed by insertion of the actual signal into the logic (e.g., rod block interlocks).

For others, there is no reason why an actual signal would preclude satisfactory performance of the test. Use of an actual signal instead of the existing requirement which limits use to a simulated signal, will not affect the performance of the channel. Operability can be adequately demonstrated in either case since the channel itself can not discriminate between " actual" or

" simul ated. "

2.

As provided for with analog channels, the signal used to test bistable channels is proposed to be allowed to be injected "as close to the sensor as practicable."

Injecting a signal at the sensor would in some cases involve.significantly increased probabilities of initiating

l*

9 undesired circuits during the test since several logic channels are often associated with a particular sensor. Performing the test by injection of a signal at the sensor requires jumpering of the other logic channels to prevent their initiation during the test,.or increases the scope of the test to include multiple tests-of the other logic channels.

Either method significantly increases the difficulty of performing the surveillance. Allowing initiation of the signal close to the sensor provides a complete test of the logic channel while significantly reducing this probability of undesired initiation.

3.

A change is proposed to allow the physical removal of a control rod to not be considered a Core Alteration.

In this activity the control cell mtst first have all the fuel bundles ren.oved prior to this control rod nvenent.

In this configuration, the negative reactivity inserted by removing the adjacent four fuel assemblies is significantly more than any minimal positive reactivity inserted during the removal of the control rod. Appropriate technical specification controls are applied during the fael movements preceding the control rod removal to protect against or mitigate a' reactivity excursion event. After such time, sufficient margin and design features (the design of a control rod precludes its removal without all fuel assemblies in the cell removed) are in place to allow removing the.TS controls during the control rod removal.

4.

Existing TS definition of LOGIC SYSTEM FUNCTIONAL TEST (LSFT) is proposed to be modified to not include the actuated device. The actuated device is proposed to be included as part of the system functional test.

In instances where the existing TS do not contain a corresponding " system functional test" which would test the actuated device, the BWR/6 Standard Technical Specifications, NUREG-1434, have proposed one be added. As an example, SR 3.6.1.6.1 is one such added system functional. test. Since the relief valve solenoid is the point where the logic and the mechanical portion of the " function" overlap, both solenoids can be tested as part of the instrumentation logic, without actuating the valve. Separately, the valve can be shown to function by actuating with either solenoid.

For completeness, however, both~ solenoids will be required to be tested in the course of two outages - as represented by the Staggered Test Basis requiremut of the Frequency.

The above less restrictive requirements have been reviewed by the staff and have been found to be acceptable, because they do not present a significant safety question in the operation of the plant.

The TS requirements that remain are consistent with current licensing practices, operating experience and plant accident and transient analyses, and provide reasonable assurance that the public health and safety will be protected.

All other definitions in the existing GGNS TS (1.2,1.8,1.10,1.14,1.15, 1.16, 1.23, 1.24, 1.26, 1.28, 1.29, 1.31, 1.32,.1.33, 1.36, 1.37, 1.38, 1.40, i

1.46, 1.48 and Table 1.1) are no longer used as defined terms in the GGNS improved TS. Definition 1.16 and Table 1.1 have been reformatted and these concepts are contained in the GGNS improved TS in Sections 1.4.

In addition,

-. - -l

e 10 definition 1.26 and 1.32 have been reformatted and these concepts are contained in the GGNS improved TS in Sections 5.0.

The remaining definitions are inapplicable under the improved TS and therefore may be deleted from the i

improved TS.

As noted above, the staff and the licensee have agreed to minor word changes throughout the GGNS improved TS 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 perform a supporting function for other sections in the GGNS improved TS.

The staff has reviewed the proposed changes in the definition section for their effect on the Safety Limits (SLs) 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 1

are not changed and, therefore, the safety margins'are not affected.

In i

1 addition, the staff concludes that the licensee's proposed changes clarify the i

definitions and would reduce the tendency for misinterpretation.

Further, the l

staff finds that GGNS improved TS definitions have appropriately applied the guidance provided in NOREG-1434. Therefore, we find the changes acceptable.

I j

1.2 Logical Connectors This is a new section in the GGNS improved TS. This section explains the meaning and use of " Logical Connectors" through the use of examples so that the entire GGNS improved TS are clearer from a human factors standpoint. We i

have reviewed this section and consider this proposed addition and reformatting to be an enhancement to the GGNS improved TS. We further find the addition to be consistent with NUREG-1434 and is acceptable.

1.3 Completion Times This is a new section in the GGNS improved TS. This section does not change i

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, 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, j

find it is consistent with NUREG-1434 and is acceptable.

I 1.4 Frequency This is a new section in the GGNS improved TS. This section defines the i

proper use and application of surveillance frequency practices through the use 1

of examples. A clear understanding of the correct application of a specified frequency is necessary to ensure compliance with a surveillance requirement.

We have reviewed this section and find that the " Frequency Notation" definition and the " Frequency Notation Table" (Definition 1.17 and Table 1.1, respectively) of the existing TS have been adequately incorporated into the

O e

i 11 descriptions and examples of this section. We find that this section is consistent with NUREG-1434 and is acceptable.

l 2.0 Safety Limits i

1 A.

Significant Administrative Changes In accordance with the guidance in the Final Policy Statement, the licensee has proposed administrative changes to the existing technical specification (TS) to bring them into conformance with the improved TS. These changes are as follows:

1.

The Safety Limits Section has been reformatted and reorganized to j

separate the safety limits and the safety limit violations. -The staff has reviewed the licensee's proposed Section 2.0, based on NL' REG-1434, i

as modified to include plant specific limits and terminology, and finds this section is consistent with the Commission's regulations and is acceptable.

The above changes result in the same limits as the current requirements, or they represent enhanced presentation of the existing TS intent. Accordingly, the improved TS changes are purely administrative and they are acceptable.

B.

Relocated Requirements In accordance with the guidance in NUREG-1434, the licensee has proposed to relocate all or portions of the existing TS 2.2, " Limiting Safety System Settings" to specification 3.3.1.1, " Reactor Protection System Instrumentation" to bring it into conformance with the improved TS.

Existing TS 2.2 specifies the reactor protection system instrumentation (RPS) setpoints and allowable values. The improved TS address these items within specification 3.3.1.1.

The protection and monitoring functions of the RPS have been designed to ensure safe operation of the reactor. This is achieved by specifying Limiting Safety System Settings (LSSS) in terms of parameters directly monitored by the RPS, as well as limiting conditions of operation (LCOs) on other reactor system parameters, and equipment performance. The LSSS are defined in this Specification as the Allowable Values, which, in conjunction with the LCOs, establish the threshold for protective system action to prevent exceeding acceptable limits, including SLs, during Design Basis Accidents. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the I

actual setpoints do not exceed the Allowable Value between successive

(

channel calibrations. Operation with a trip setpoint less conservative I

than the nominal trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. Thus, the RPS setpoints are l

effectively retained within the improved TS. This change is considered I

an administrative change in the location of the requirements within the j

TS and is therefore acceptable.

j l

l i

l

-O 9.

12 The licensee has proposed to relocate to the UFSAR or Technical Specification Bases a portion of existing TS 2.1.4 which describes operator action to manually initiate the emergency core cooling system (ECCS) to restore reactor vessel water level, after depressurizing the-reactor vessel, if required. The improved TS 2.2.2 Required Action has' been made less specific by allowing operator flexibility in determining the best method to restore the water level. Directions for the methods to be used for compliance depend on plant conditions. The relocated-operator instruction does not affect plant configuration or change existing plant operating practices.

The improved TS_still require the manual backup capability to the automatic plant features that provide ECCS system injection. Therefore, this requirement ~can.be relocated to the UFSAR without affecting plant safety.

In addition, the time-frame for completion of the action is made consistent with the allowed time to-restore other SL violations.

The above. relocated requirements relating to reactor vessel water level are not required to be in the TS under 10 CFR 50.36, and are not required to obviate the possibility of an abnormal situation or event giving rise to an immediate threat-to the public health and safety.

Further, they do not fall within any of the four criteria set forth in the Commission's Final Policy.

Statement,. discussed in the Introduction above.

In addition, the staff finds that sufficient regulatory controls exist under 10 CFR 50.59. Accordingly, the staff has concluded that these requirements may be relocated from the TS to the UFSAR or TS Bases, as applicable.

C.

More Restrictive Requirements-By electing to implement NUREG-1434, Section 2.0, Safety Limits, the licensee has proposed a number of more restrictive conditions than are required by the existing TS. These conditions are the following:

1.

The applicability of existing TS 2.1.1, 2.1.2, 2.1.3, and 2.1.4 is extended to all Modes of operation. Although it is physically impossible to violate some SLs in some Modes, any SL violation will receive the same attention and response.

2.

Safety Limit 2.1.2 and 2.1.3 limits on steam dome pressure and core flow are to be specified as " equal to or greater than". The current SLs do not address a pressure or flow which is equal to the limit. This proposed change will resolve an inconsistency between current SL 2.1.1 and SL 2.1.2.

The staff has reviewed the above more restrictive generic requirements and concludes that they result in an enhancement to the improved TS. Therefore, the more restrictive requirements are acceptable.

3.0 Limiting Conditions for Operation This section has been renamed from the existing technical specification (TS) section " Limiting Conditions for Operation and Surveillance Requirements" to

o o

l i

13 4

the GGNS improved TS section entitled " Limiting' Condition for Operation (LCO)

Applicability" and " Surveillance Requirement (SR) Applicability." The following covers changes made throughout Section 3.0.

The licensee has proposed to relocate within the TS administrative controls the remaining portions of existing TS Surveillance Requirement 4.0.5 that require American Society of Mechanical Engineers (ASME) i inservice testing:

4.0.5.b, surveillance intervals specified in ASME Section XI; 4.0.5.c, limitations to extending required surveillance intervals; 4.0.5.d,- requirements to perform ASME tests in addition to i

other TS required surveillances; and 4.0.5.e, requirements that ASME i

tests do not supersede any other TS.

i i

This change is considered an administrative change in the location of the-j requirements within TS, and is therefore considered acceptable.

In accordance with the guidance in the Final Policy Statement the i

licensee has 'also proposed to relocate to the UFSAR a portion of existing TS Surveillance Requirement 4.0.5.a that specifies requirements j

for inservice inspection of ASME Code Class 1, 2, and 3 components out i

of the GGNS improved TS. The requirements state that inspections shall be performed in accordance with Section XI of the ASME Boiler & Pressure Vessel Code and applicable Addenda. The same requirements'are specified j

by 10 CFR 50.55a(g) unless specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).

The staff concludes a

that control of this commitment under 10 CFR 50.55a is' acceptable and i

that the regulatory requirements need not be restated in the TS.

l The licensee has also proposed to add three new LCOs from NUREG-1434 (LC0 3.0.5, LCO 3.0.6 and LCO 3.0.7) to the GGNS improved.TS.

1.

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

i 2.

LCO 3.0.6 is being added to permit an exception to LCO 3.0.2.

3. -

LCO 3.0.7 is being added to permit performance of special tests and operations.

4 In addition, clarifying statements have been added to SR 3.0.2, SR 3.0.3, SR 3.0.4 and LCO 3.0.4.

1.

SR 3.0.2 is clarified to provide a completion time extension for each performance of a periodic surveillance requirement.

2.

SR 3.0.3 quantifies and clarifies the maximum time delay or allowance that is permitted to perform a given surveillance.

I' 3.

SR 3.0.4 and LC0 3.0.4 clarify limitations on mode applicability changes during shutdown conditions and power reductions.

4 i

s

,m.

.-,n,

-.,,,.,w

i -;.

14 i

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

The staff concludes that the proposed GGNS improved TS have made appropriate i

application of the guidance provided in NUREG-1434 and finds the changes are i

l acceptable.

l l

3.1 Reactivity Control Systems i

A.

Significant Administrative Changes.

i In accordance with the guidance in the Final Policy Statement, the licensee

(

has proposed administrative changes to the existing technical specifications (TS) to bring them into conformance with the improved TS. These' changes are i-as follows:

l 1.

. Existing.TS 3/4.1.1 had the following administrative changes:

a.

In Modes 3 and 4, a single control rod may be withdrawn under. the provisions of the improved TS 3.10.3 and TS 3.10.4, or some unanticipated event may have resulted in control rods not being inserted. Therefore, rather than the existing TS 3.1.1 Action b.

requirement to " verify all insertable control rods to be inserted,"

the improved TS 3.1.1 Required Actions-(ras) C.1 and D.1 require, j

" Initiate action to fully insert all insertable control rods."

l This wording provides the same intent in the event all insertable i

control rods are not found to be inserted, but also clarifies that any control rods not inserted are to be inserted, b.

In Modes 3 and 4, the vessel head is bolted in-place and the only activity that could significantly reduce Shutdown Margin (SDM) is control rod withdrawal. Since improved TS 3.1.1 ras C.1 and D.1 i

ensure control rods remain inserted, any additional action to suspend activities that could reduce the shutdown margin is redundant and unnecessary. Similarly, in Mode 5 the only activities that could-affect shutdown margin are core alterations and control rod withdrawal. Since improved TS 3.l~1 ras E.1 and E.2 require j

suspension of core alterations and ensure that control rods remain j

inserted, any additional action to suspend other activities is-redundant and unnecessary.

i c.

The existing TS Actions b. and c. requirements to establish secondary i

containment integrity within eight hours would appear to provide eight hours in which the integrity could be violated even if capable of being maintained. Additionally, if the plant status is such that integrity is not capable of being established within eight hours, the-existing TS Action results in non-compliance with the TS and a requirement for an Licensee Event Report (LER).

Improved TS 3.1.1 ras more appropriately present the intent of the existing TS Actions D.2, D.3, and D.4, and E.3, E.4, and E.5 with a requirement to establish and maintain the secondary containment boundary within one 4

i

-w

-m-gn

i i

L j

15 hour1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />. No longer would the provision exist which seem to allow a j

violation of the boundary for up to eight hours.

l d.

Improved TS 3.1.1 RA E.2 revises the existing TS Actions c.

i requirement from insert all insertable control rods within one hour i

to initiate action to fully insert all insertable control rods 4

immediately. The existing TS requirement would appear to provide j

7 one hour in which control rods could be left withdrawn, even if i

insertable. Also, if the control rod is incapable of being inserted in one hour, the existing TS action would appear. to result in the j

i requirement for an LER.

Improved TS RA E.2 more appropriately.

j presents the intent of the existing TS action. This RA imposes a more conservative requirement to insert and maintain inserted the d

I control rod (Q. No longer would the provision appear to allow the j

operators to withdraw or leave withdrawn one or more control rods for up to one hour.

e.

Improved TS Surveillince Requirement (SR) 3.1.1.1 includes a specific j

performance. time for.he existing TS 4.1.1 a. SDM test, to clarify R

i when, " prior to or during the first startup," it.is to be' performed.

J Interpretations of the time available to perform the SDM tests, both i

more and less conservative, could be made for the existing TS:

requirement. Most SDM tests are in-sequence control rod withdrawal 3

criticality tests. Therefore, the improved TS provides four hours after reaching criticality-as a reasonable time to perform the i

calculations and complete verification.

3 l

l j

f.

The existing TS SR 4.1.1 a. wording "after each refueling" does not' i

convey the intent of the activity to perform the SDM test after 1

in-vessel activities that can alter the SDN.

Improved TS SR 3.1.1.1 more explicitly presents these activities as " fuel movement within l

the reactor pressure vessel or control rod replacement."

g.

Improved TS 3.1.1 ras D.2, D.3, and D.4,.and E.3, E.4, and E.5.

replace the use in existing'TS LCO 3.1.1 Action b. of the defined term SECONDARY CONTAINMENT INTEGRITY with the essential elements of-j that definition. The change is editorial in that the improved TS ras j

specifically address all the requirements.

1 l

2.

Existing TS 3/4.1.2 had the following administrative change:

Improved TS SR 3.1.2.1 includes a specific time for completing the reactivity anomaly surveillance. This modification clarifies the i

existing TS 4.1.2 a. requirement as to when "during the first startup" the test must be performed. This test is performed by comparing the k,,,

i from the core performance log to the vendor provided target k,,, as a d

function of cycle exposure while at steady state reactor power l

conditions. The improved TS provide 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching these conditions as a reasonable time to complete the calculations and obtain 4

appropriate verification.

h j

4-

j.

5 16-i 3.

. Existing TS 3/4.1.3.1 had the following administrative changes:

a.

Improved TS 3.1.3 reorganizes and simplifies existing TS 3/4.1.3.1 to 7

include all existing TS conditions that can affect the ability of the 4 '

control rods to provide the necessary reactivity control as follows:

(1) A control rod is considered inoperable only when it is degraded

.such that it cannot provide its scram function; an' inoperable.

J control rod (except a stuck rod) must be fully inserted and l

disarmed; (2) "untrippable" is not treated as " stuck" provided the control" rod l

l can be fully inserted; if only the "trippable" capability is lost, the control rod is considered inoperable and must be'

+

i inserted and disarmed-I (3) a control rod is considered " inoperable" and " stuck" if it is i

incapable of being inserted-I (4) a control rod is considered " slow" when it is capable' of providing the scram function but is not able to meet the scram time limits; and (5)- special considerations provide conformance to the Banked i

Position Withdrawal Sequence (BPWS) at less than 10% of rated 1-thermal power.

t The scram reactivity in the safety. analysis specifies'the number of inoperable and slow scramming rods, and the control rod drop accident j

analysis provides considerations of the BPWS at low power levels.

}

i b.

Improved TS 3.1.3 and 3.1.8 Actions Notes-provide more explicit j

instructions for proper application of the actions for improved TS l

compliance. With improved TS 1.3, " Completion Times," these notes provide direction consistent with the intent of the existing TS l

actions for inoperable control rods (improved TS 3.1.3) and Scram i

Discharge Volume (SDV) vent and drain valves (improved TS 3.1.8).

In i

the first case, the improved TS allow a specified period of time to j

verify each inoperable control rod is in compliance with certain

{

limits and, when necessary, insert and disarm the control rod.

[

Additionally, the improved TS allow a specified period of time.to confirm that each SDV line is isolated or capable of isolation, and e

{

to restore the complete design function of the line.

c.

Improved TS 3.1.3 replaces the existing TS's word " immovable," as a i

result of excessive friction or mechanical interference or known to be untrippable, with the term " stuck." The improved TS 1

simplification retains the intent of the existing TS wording.

1 Details of potential mechanisms, mechanical or electrical, by which control rods may be " stuck" are not necessary in the action.

i i

I r

a

L.

1 17 1'

d.

Due to the improved TS changes to the requirements for inoperable control rods, there are no withdrawn operable control rods required to have directional control valves disarmed. Existing TS 4.0.3 and.

improved TS SR 3.0.1 do not require inoperable control rods to meet 1

L this surveillance, and the improved TS do not require operable control rods to have their directional control. valves disarmed.

Therefore, this change is a deletion of an unusable allowance.

4 e.

In a letter dated April 21, 1994 (GNRO-93/00036), the licensee proposed changes to the frequency for demonstrating control rod-

. operability above the low power setpoint of the rod pattern control.

system by changing rod positions one notch from at least once per 7.

days for all withdrawn rods to once per 7 days for fully withdrawn-l l

rods and once per 31 days for partially withdrawn rods. The staff j

approved this change in Amendment No.ll5, dated February 16, 1995.

l 1

Existing TS 3/4.1.3.2 had the following administrative changes:

4.

i Improved TS 3.1.4 reorganizes the control rod scram insertion time j

specification as follows i

4 (1) It relocates to the LCD statement specified conditions that allow unrestricted operation with slow control rods. This change states in the LCO that operation is allowed to continue without requiring a i

j label of also having an LCO not met.

j (2) It eliminates the category of " fast" control rods and any requirement on average scram times, thus simplifying the presentation.

In the improved TS, the LCO limit is the existing TS " fast" average limit.

Changes in the allowed number of " slow" control rods and in elimination of any average scram time requirement offset the

)

1 additional restriction on the LC0 scram time. These changes are 1

i specifically the subject of Less Restrictive comment 4.a. in l

subsection D.

l (3) It presents the LCO scram time limits in Table 3.1.4-1.

i j

5.

Existing TS 3/4.1.3.3 had the following administrative changes:

1

\\

a.

In general, improved TS 3.1.5 changes the outline for the format and editorial contents of the existing TS by rewriting the existing TS Actions a.1 and a.2 as Conditions A, B, and C:

(1) The improved TS changes existing TS Action a.1 for one inoperable accumulator at any reactor pressure into Conditions A and C.

Condition A is for one inoperable accumulator at reactor pressures that will support control rod insertion, and Condition C is for reactor pressures that will not support rod insertion; j

and 1-a i

2.

i

1

'O 4-l.i i

i 18 l

1 (2)' The improved TS splits existing TS Action a.2 for two or'more inoperable accumulators into Conditions B and C, dependent on l

reactor pressure.

t i

l b.

The improved TS includes the Note, " Separate condition entry is allowed for each control rod scram accumulator,".to provide more explicit instructions for proper application of the actions for TS l

l compliance. With improved TS 1.3, " Completion Times," this Note i

gives direction consistent with the intent of.the existing TS actions-for inoperable control rod accumulators. Upon discovery of each inoperable accumulator,-the improved TS intends that each specified i

action be applied regardless of it having been applied previously for i

j other inoperable accumulators.

l 3

c.

The improved TS actions for inoperable accumulators do not repeat the action to insert and disarm an inoperable control rod or the shutdown requirement'for failure to perform this action. The existing TS

}

Action a.2 requires the affected control rod to be declared j

inoperable. Once declared inoperable, the improved TS 3.1.3 Actions l

for an inoperable control rod are applied, requiring insertion and j

disarming.

d.

The improved TS maintains the requirement to verify that a control rod drive pump is operating, but changes the method of verifying this i

from inserting the control rod one notch to verifying that the

[

charging water header pressure is at least:1520 psig.

Either method ensure that sufficient control rod drive pressure exists to insert i

control rods. The improved TS method of determining charging water header pressure provides added assurance that the charging header pressure is sufficient to drive all rods, whereas the existing TS

-method only assures that one rod is still capable of insertion.

Since the change is merely exchanging one test method'for another equivalent or better test method, this change is considered administrative.

6.

Existing TS 3/4.1.3.4 had the following administrative changes:

a.

Improved TS SR 3.1.3.5 presents the requirement that control rods be i

coupled to their drive mechanisms, making it a requirement for control rods to be considered operable. The actions for uncoupled control rods remain effective.

Eliminating the separate existing TS

]

LCO for control rod coupling, by moving the surveillance and actions to another specification, does not eliminate any requirements, or impose a new or different treatment of the requirements.

j b.

The existing TS does not clearly present the tioe allowed for the action to insert and disarm the control rod. The action follows an action to declare the control rod inoperable.

It is-interpreted to give direction as to the proper option for inoperable control rod actions, similar to existing TS 3.1.3.1, Action b.1, with two options.. The existing TS 3.1.3.1 Actions for inoperable control rods allow one hour from the time the control rod is declared inoperable l

i

~.. -

- ~.

i.

19 y

1 until it is required to be inserted and disarmed. Therefore, the j

intent of-the existing TS requirement is deemed to allow a total of i

three hours (two hours in existing TS 3.1.3.4,' Action a., plus the i

one hour discussed here). This total of three hours is consistent with the improved TS ras for inoperable control' rods and it applies 4

i to uncoupled control rods as well.

4 c.

Existing TS 4.1.3.4.a b and c address the requirement to perform control rod coupling checks after performing activities that could have affected control rod coupling integrity. This surveillance must be completed before declaring the control rod operable. The rewrite of existing TS 4.1.3.4.c into improved TS SR 3.1.3.5 more clearly i

presents control rod operability considerations.

In improved TS SR 3.1.3.5, core alterations that could have affected control rod i

drive coupling integrity are considered to be a subs of the i

3 existing TS 4.1.3.4 requirement to perform operabilit. tests i

following maintenance which could have affected the cowrol rod drive coupling integrity. Thus, combining existing TS 4.1.3.4.a,-b and c into a single surveillance requirement represents no actual change in requirements.

1 7.

Existing TS 3/4.1.3.5 had the following administrative changes:

Improved TS SR 3.1.3.1 includes the interpretation of the existing TS i

3.1.3.5 requirement that a control rod is operable when the. position of each control rod is determined. The existing TS 3.1.3.5 requires at least i

one control rod position indication system.

In the existing TS, the LCO requirement, Applicability footnote *, Action a., and SR 4.1.3.5 specify J

requirements for individual control rods. The-instrumentation LCOs for existing TS 3.3.6 and improved TS 3.3.2.1 for the RPCS and the RWL

'j contain system operability requirements that require proper control rod position indication input.

j' Improved TS SR 3.1.3.1 requires that the control rod position be.

determined, which requirement meets the. intent of both the existing TS 8

LCO and Action a.l.

If-the position 'can be determined, the control rod may be considered operable, and continued operation may be allowed.

Meeting improved TS SR 3.1.3.1 has the same.effect as complying with existing TS Action a.1.

Therefore, this change is administrative.

8.

Existing TS 3/4.1.4.2 had the following administrative change:

l The improved TS move the existing TS requirements for the rod pattern control system (RPCS) to the Instrumentation Section (improved TS

)

i 3.3.2.1). However, improved TS 3.1.3 still addresses the parts of existing TS 3/4.1.4.2 that contain requirements and actions not directly associated with the operability of the RPCS. They include:

4

1) determination of SDM with a stuck control rod, existing TS Action b.1;

2) positioning and disarming of inoperable control rods, existing TS Action b.2.a; 3) inoperable control rod separation requirements, existing TS Action b.2.b; and 4) control rod pattern out-of-sequence constraints, existing TS Action b.3.

I 1:

\\

_.,._,,y

20 9.

Existing TS 3/4.1.5 had the following administrative change:

Improved TS 3.1.7 clarifies existing TS 4.1.5 b.2 Footnote

• and d.2 Footnote **, which require these surveillances be performed when the Standby Liquid Control System solution temperature drops below the specified limit. With the temperature below this limit the system is i

inoperable. The required verificatinn is not productive during the time the temperature is low. The improved TS clarifies the intent of this requirement is to confirm, after restoring the temperature and therefore 1

the system to operable status, that the low temperature condition did not result in precipitation.

J The above changes result in the same limits as the current requirements, or they represent enhanced presentation of the existing TS intent. Accordingly, the improved TS changes are purely administrative and they are acceptable.

B.

Relocated Requirements In accordance with the guidance in NUREG-1434, the licensee proposed to relocate all or portions of the following existing TS within the improved TS:

Existina TS Title I

3/4.1.3.3 Control Rod Scram Accumulators 3/4.1.3.5 Control Rod Position Indication 3/4.1.4.1 Control Rod Withdrawal 3/4.1.4.2 Rod Pattern Control System The more significant changes resulting from relocated items are as follows:

1.

Existing TS 3.1.3.3 Action b. and Footnote

• specify the action requirements for one or more inoperable control rod scram accumulators in Operational Condition 5.

The improved TS address these requirements in Specification 3.9.5, Control Rod Operability - Refueling.

Thus, the improved TS effectively retain the action requirements for one or more l

inoperable control rod scram accumulators in Mode 5.

l 2.

Existing TS 3.1.3.5 Action a.3.a)2) requires the verification of position l

and bypassing of control rods with inoperable Full-in and/or Full-out position indicators, when below the low power setpoint of the RPCS. The improved TS address this requirement in improved TS 3.3.2.1, control Rod i

Block Instrumentation. Therefore, the improved TS effectively retain the

)

requirement to verify the position and bypassing of control rods with inoperable Full-in and/or Full-out position indicators.

l 3.

Existing TS 3.1.3.5 Action b. and Footnote

• specify the action requirements for both control rod position indicators for a withdrawn control rod inoperable in Operational Condition 5.

The improved TS address these requirements in specification 3.9.4, " Control Rod Position Indication." Thus, the improved TS effectively retain the action requirements for both control rod position indicators for a withdrawn control rod inoperable in Mode 5.

e

v i

21 4.

' Existing TS 3/4.1.4.1 that specifies requirements for con.rol rod withdrawal and the portion of 3/4.1.4.2 that specifies requirements for i

the RPCS are control rod program control requirements that are relocated to Specification 3.3.2.1, Control Rod Block Instrumentation. The H

improved TS address each of the requirements in these LCOs, except for j

existing TS 3.1.4.2 Actions for inoperable control rods which are retained in Specification 3.1.3, Control Rod Operability. Thus, the i

j improved TS effectively retain the requirements for both the control rod withdrawal and for the RPCS.

i i

5.

Existing TS SR 4.1.5.d.2 verification of the relief valve is proper i

operation and setpoint is conducted in accordance with the plant's Inservice Test Program referenced in improved TS LCO 5.5.6 and the ASME j

code.

]

The above changes are considered purely administrative changes in the location of the requirements in the improved TS and are therefore acceptable.

In accordance with the guidance in the Final Policy Statement, the licensee has proposed to relocate all or part of the following existing TS to other licensee-controlled documents:

Existina TS I_111g 3/4.1.2 Reactivity Anomalies 3/4.1.3.1 Control Rod Operability 2

3/4.1.3.2 Control Rod Maximum Scram Insertion Times J

3/4.1.3.3 Control Rod Scram Accumulators i

3/4.1.3.4 Control Rod Drive Coupling i

3/4.1.3.5 Control Rod Position Indication i

3/4.1.3.6 Control Rod Drive Housing Supports j

3/4.1.4.2 Rod Pattern Control System j

3/4.1.5 Standby Liquid Control System l

1.

Existing TS 3.1.2 Action a. includes a requirement' to re-evaluate j

j predicted core reactivity conditions to explain the cause and correct the difference in predicted values, so that, based on the new' evaluation,.the reactivity difference is returned to within' acceptable limits..The 4

i relocation of the instruction specifying that an analysis is needed to explain the cause of core reactivity conditions that depend on plant i

j conditions can be appropriately controlled in the Bases and is made to be consistent with the NUREG-1434 presentation, j

2.

Existing TS 3.1.3.1, Actions a.1.b), b.1.), and b.2; existing TS 3.1.3.5, Action a.3.b); and existing TS 3.1.4.2, Action b.2.a), present specific i

alternative methods for disariting an inoperable control rod's drive. The improved TS do not contain procedural details of how to perform a l

specific action. Specifying the procedural details of alternative l

methods for disarming an inoperable control rod's drive depends on plant conditions and operator actions and can be appropriately controlled in the Bases.

Further this change is consistent with the NUREG-1434 j

presentation.

l Y

22 3.

Existing TS 4.1.3.2.c requires a representative sample of control rods tested each 120 days of power operation, and the existing TS Action c.1 imposes a limit on the number of slow control rods in this sample.

They delineate the scope of this sample and the statistical limit for slow control rods in the sample. Specifying the scope of this sample and the statistical limit for slow con',rol rods in the sample depends on plant conditions and operatar actions and can be appropriately controlled outside of the TS.

Furthr., this change is made to be consistent with the NUREG-1434 presentation.

4.

Existing TS 4.1.3.3 b. states the surveillance requirements for scram accumulator leak detectors, pressure detectors, and associated alarm instrumentation. The improved TS does not require indication-only or test equipment to be operable to support the operability of a system or component. The scram accumulator leak detectors, pressure detectors, and associated alarms do not relate directly to accumulator operability.

Further, this change is made to be consistent with the NUREG-1434 presentation. The requirements have been relocated to the UFSAR and Bases.

5.

Existing TS 3.1.3.4 Actions a.1. and a.2, spell out the requirements for one control rod not coupled to its associated drive mechanism. Action a.1 contains detailed methods of restoring coupling integrity to an uncoupled control rod. The improved TS actions do not explicitly detail options on how to restore the control rod to operable status. This action is always an option, and is implied in all conditions. Thus, omitting this action is purely editorial. The Bases present the details contained in the existing TS actions.

Furthermore, the requirement of improved TS SR 3.0.1 on improved TS SR 3.1.3.5 addresses the specific detailed action requiring the control rod to be declared inoperable (existing TS 3.1.3.4, Action a.2).

SR 3.0.1 requires that failure to meet a Surveillance shall be failure to meet the LCO.

In the improved TS, failure to mut SR 3.1.3.5 for a control rod results in the control rod being considered inoperable. Specifying the options on how to restore the control rod to operable status depends on plant conditions and operator actions and can be appropriately controlled outside of the TS. Further, this change is made to be consistent with the NUREG-1434 presentation.

6.

Existing TS 3.1.3.5 Actions a.1 and a.2' require the determination of the position of the control rod. The action can be met two ways: by reference to the alternate control rod position indicator, and by moving the control rod to a position with an operable position indicator. The improved TS do not contain procedural details of how to perform a specific action. Specifying how to determine the position of a control rod depends on plant conditions and operator actions and can be appropriately controlled outside of the TS.

Further, this change is made to be consistent with the NUREG-1434 presentation.

7.

Existing TS 4.1.3.5 b. provides detailed methods to verify control rod position already found in existing procedures for the performance of the surveillances referenced therein. To perform control rod movement tests

L.

23 (existing TS 4.1.3.1.2) and coupling verification (existing TS 4.1.3.4) i requires position indication availability.

If position indication is not

- available, these tests cannot be satisfied and appropriate actions will be taken for inoperable control rods.

If at'any time the position of a l

control rod is unknown, the control rod would be considered inoperable,

)

and appropriate actions taken. The improved TS'do not contain procedJral l

details of how to perform a specific surveillance. Specifying the-options on how to restore the control rod to operable status depends on i

plant conditions and operator actions and can be appropriately controlled outside of the TS. Further, this change is made to be consistent with the NUREG-1434 presentation.

8.

Existing TS 3.1.3.6, " Control Rod Drive Housing Support" that supports control rod operability by plant configuration management is relocated to the UFSAR. As such, control rod operability (improved TS 3.1.3) cannot be satisfied without the support being in place. Without control rod operability confirmed, appropriate actions of the control rod operability specification must be entered. There is no need for duplicate l

requirements in a subsystem LCO.

Relocation of this LCO is appropriate since plant configuration (the control rod housing support in place) would be controlled by post-maintenance procedures.-

9.

Existing TS SR 4.1.3.1.4.b duplicates existing TS SR 4.3.1.1 for the testing of the SDV level scram instrumentation. The improved TS moves SR 4.1.3.1.4.b for the scram function to the RPS TS (3.3.1.1) and relocates the control rod block function of the scram discharge level i

instrumentation to the control of 10 CFR 50.59.

10. Existing TS 3.1.3.3 Actions c. and d. delineate the requirements when one or more control rod scram accumulator pressure detectors or alarms, or l

leak detectors or alarms, are inoperable. The improved TS does not require indication-only or test equipment to be operable to support the operability of a system or component. The scram accumulator leak detectors, pressure detectors, and associated alarms do not relate directly to accumulator operability. Therefore, these requirements do not satisfy the Policy Statement for inclusion in the improved TS.

11. The improved TS delete existing TS SR 4.1.3.3 b.2 requirements for taking measurements to trend the ability of accumulator check valves to maintain accumulator pressure. No accident or transient analysis assumes the control rod scram accumulator check valves maintain accumulator pressure for a specified time if no control rod drive pump is operating. With no operating pump, improved TS 3.1.5 ras B.1 and D.1 require the reactor-scrammed if one pump is not restored to operation within 20 minutes after two or more accumulators have low pressure.

12.

Improved TS 3.1.6 RA A.1 requires restoration of control rods to their proper in-sequence positions, in compliance with BPWS, without specific direction as to how or in what order, this must be performed. The existing TS 3.1.4.2 Action b.3.f) provides specific rod movement requirements that.are not necessarily evident in practice. Control rod position is related to the other control rods in its group when

_ ~.

)

i 24 determining its in-sequence status. A control rod withdrawn _beyond i

another control rod in its group could be considered withdrawn past its j

i in-sequence position. However, the situation could be viewed as the other control rod (s) being. inserted beyond their in-sequence position (s).

j The detail and complexity on recovery processes when control rod (s) are l

discovered out of sequence are relocated to the TS Bases.

a

13. Existing TS 3/4.1.3 specifies in many of the specifications the details j

of the methods for performing certain surveillance requirements. The i

details of the methods and acceptance values for these tests are located in and adequately controlled by plant procedures and improved TS Bases.

i The values are system design values which were previously reviewed and 4

approved by the staff, and are also controlled by the design change

{

procedures and 10 CFR 50.59.-

)

14. Existing TS 3/4.1.5 specifies standby liquid control system (SLCS) i requirements to perform testing during operation and during shutdown of l

l the SLSC pump, flow path and replacement charges. The method of i

j-performing surveillance testing is relocated to plant procedures. The requirement that replacement charge and pump relief valve testing be.-

j performed "during shutdown" is relocated to procedural, administrative i

controls in accordance with the guidance of Generic Letter. 91-04..

i Requirements on the replacement charges for explosive ~ valves have been relocated to the Bases and plant administrative controls. Verification of the relief valve proper operation and setpoint is conducted in 4

)

accordance with the plant's Inservice Test Program and the ASME code.

The design features and system operation which dictate the methods 1

identified in the existing TS are described in the UFSAR.

Since the i

procedural details of how a specific surveillance is performed are not to j

be located in the improved TS under NUREG-1434 and are not relied upon to i

prevent or mitigate the consequences of a design basis accident, the i

methods can be relocated to the Bases-and removed from the TS.

The above relocated requirements relating to reactivity control systems are j

not required to be in the TS under 10 CFR 50.36,-and are not required to j

obviate the possibility of an abnormal situation or event giving rise to an j

immediate threat to the public health and safety.

Further, they do not fall within any of the four criteria set forth in the Commission's Final Policy Statement, discussed in the Introduction above.

In addition, the staff finds l

that sufficient regulatory controls exist under 10 CFR 50.59. Accordingly, the staff has concluded that these requirements may be relocated-from the TS i

l

.to the licensee's TS Bases or to the UFSAR as applicable.

)r C.

More Restrictive Requirements.

1 By electing to implement the NUREG-1434 Section 3.1 Specifications, the i

licensee has adopted a number of more restrictive conditions than are required q

by the existing TS. The more significant conditions are the following:

d

{

1.

The improved TS add another surveillance frequency for SDM verification than is specified in existing TS 4.1.1.

This new requirement clarifies the need for ensuring SDM during the refueling process.

Because many i

e

)

i 1

4

,e

[

25 o

i refueling mode analyses in the UFSAR require adequate SDM, measures must

)

ensure SDM for the intermediate fuel loading patterns during refueling.-

This change imposes a requirement where none is explicitly provided in the existing TS. This new requirement does not, however, require 1

introducing tests or modes of operation of a new or different nature, j

i As presented in the improved TS Bases for this new requirement, an l

analysis best accomplishes ensuring SDM for refueling because of the many l

j changes in the core loading during a_ typical refueling.

Bounding j

analyses may be used to demonstrate adequate SDM for the most reactive configurations during refueling, thereby showing acceptability of the entire fuel movement sequence.

l 2.

Existing TS 3.1.3.1 Action a. would appear to require LC0 3.0.3 entry if more than one control rod were stuck. The improved TS 3.1.3 Conditions A i

and B ras maintain the equivalent shutdown action as LCO 3.0.3, but also i

contain an additional requirement to disarm the stuck control rod. This i

additional requirement provides a level of protection to the control rod i

drive should a scram signal occur.

If mechanically-bound, the stuck j

control rod could cause further damage if not disarmed.

i

)

j 3.

Improved TS 3.1.3 ras C.1 and C.2 for non-stuck inoperable. control rods replace the existing TS 3.1.3.1 Action b.1.b) and associated footnote j

check of insertion capability with requirements to fully insert and disarm all inoperable control rods. The existing TS action, requiring the insertion capability verification and then allowing the control rod to remain withdrawn, is applicable to conditions such as:

1) slow i

j control rods; 2) one inoperable control rod drive scram accumulator; and '

j

3) loss of position indication while below the low power setpoint. The first and second of these conditions are addressed later in comments for existing TS 3.1.3.2 and 3.1.3.3, respectively. The third condition would 4

j no longer allow the affected control rod to remain withdrawn and not j

disarmed. This added restriction on control rod (s) with loss of position j

indication is conservative with respect to scram time and SDM.

Improved l

TS 3.1.3 Condition D ras for inoperable control rods not complying with the BPWS ensure that insertion of these control rods remains j

appropriately controlled.

l 4.

Improved TS SRs 3.1.3.2 and 3.1.3.3 require control rods to be inserted, l

in lieu of the existing TS 4.1.3.1.2 requirement for control rod 2

movement. The existing TS requirement could be met by control rod i

withdrawal.

It is conceivable that a mechanism causing binding of the control rod that prevents insertion could exist such that a withdrawal 3

test would not detect the problem. Since the purpose of the test is to i

assure scram insertion capability, restricting the test to only allowing j

control rod insertion provides an increased likelihood of this test j

detecting a problem that impacts this capability, d

j 5.

Improved TS SRs 3.1.4.3 and 3.1.4.4 delete the flexibility provided by i

existing TS 4.1.3.2 b Footnote ** to delay post maintenance testing until prior to entry into Operational Condition 1 from the' existing TS l

1 Surveillance Requirement for specifically affected individual control rods. This change ensures adequate testing is performed before declaring

\\

j *:

b 26 the control rod operable and entering Mode 2.

In support of this restriction, improved TS 3.1.4.3 requires a scram time test, at any a

4 reactor pressure, before declaring the control rod operable and, thus, enabling its withdrawal during a startup. To allow testing at less than l

normal operating pressures, additional scram time limits will be i

established based on manufacturers' data.

If. the control rod remains 1

inoperable, then it must be inserted and disarmed until normal operating pressures are reached. Since the existing TS do not require this test, i

the improved TS limits are construed to be a more restrictive l

{

requirement.

Furthermore, the existing TS scram time test performed at i

normal operating reactor pressure is still required before exceeding 40%

i t

power. These limits are reasonable for application as a test of operability at these conditions.

6.

Improved TS 3.1.5 RA A.1 or RA A.2 allows eight hours when a single accumulator is inoperable, but only if the reactor pressure is i

sufficiently high to support control rod insertion.- Existing TS 3.1.3.3 l

Action a.1 allows eight hours for one accumulator at any reactor j

pressure. At reduced reactor pressures, a control rod may not insert on-a scram signal unless the associated accumulator is operable. The allowances in improved TS LCOs 3.1.3 and 3.1.4, for number and j

distribution of inoperable and slow control rods, do not justify an j

additional control rod failing to scram due to inoperable accumulator and i

low reactor pressure for up to eight hours without compensatory action.

1 Therefore, improved TS 3.1.5 Condition A reflects existing TS 3.1.3.3 1

Action a.1 for one inoperable accumulator during sufficiently high l

reactor pressure, and improved TS 3.1.5 Condition C reflects existing TS

^

3.1.3.3 Actions a.1 and a.2 for lower reactor pressures for one or more l

inoperable accumulators.

5 i

7.

Existing TS 3.1.3.1 Action b.,' for inoperable control rods, provides the option to verify the insertion capability, and then allows the control l

rod to remain withdrawn.

Improved TS 3.1.3 Condition C changes the 1

action for a non-stuck inoperable control rod by replacing the check of insertion capability with a requirement to fully insert and disarm each i

inoperable control rod.

Item 3 just above discusses these changes. The effect on the action of existing TS 3.1.3.5 Action a.3.a)l) for a control i-rod with its position unknown, when below the low power setpoint of the j

RPCS, is to eliminate the option to declare the rod inoperable and leave it withdrawn and continue to operate. The rod must be inserted and disarmed, regardless of the power level; the existing TS only requires j-this action if power is greater than the low power setpoint.

The staff has reviewed these more restrictive requirements and concludes.they 1

i result in enhancement to the existing TS.

Therefcre, these more restrictive i

j requirements are acceptable.

i D.

Less Restrictive Requirements.

l The licensee, in electing to implement the NUREG-1434 Section 3.1 j

Specifications, proposed a number of less restrictive conditions than are i

r i

i-h' 1

,a d

O O

1 27 1

j allowed by the existing TS. The more significant conditions are the i

following:

a 1.

Existing TS 3/4.1.1 had the following less restrictive changes:

a.

Improved TS 3.1.1 ras for Condition E change the existing TS 3.1.1 Action c. requirements in Mode 5 when SDM is less than specified,1) to add "except for control rod insertion and fuel assembly removal" to " suspend Core Alterations," and 2) to only require those control rods in core cells containing one or more fuel assemblies to be fully inserted. The first change allows the continuation of activities that can correct the problem and restore the margin of. safety.

For the second, since inserting the rod with no fuel in the cell has negligible impact on SDM, it is acceptable to. provide this flexibility.

b.

Improved TS SR 3.1.1.1 eliminates the existing TS 4.1.1 b. frequency for determining SDM equal to or greater than specified at any time during the fuel cycle. The specified SDM limits account for uncertainties, biases, and fuel cycle changes. As long as this margin is met, as determined by the initial startup test and corroborated by the periodic reactivity anomaly surveillance (existing TS 4.1.2 and improved TS 3.1.2.1), there is no need for additional determinations.

The improved TS moves the existing TS 4.1.1 c. requirement to verify

c..

SDM with a control rod stuck to improved TS 3.I'.3 RA A.3.

In doing so, improved TS 3.1.3 RA A.3 changes the existing TS 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> frequency to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. With one control rod stuck in a withdrawn position, the remaining operable control rods are capable of-providing the required scram and shutdown reactivity.

Failure to reach cold shutdown is only likely if a control rod adjacent to the stuck control rod also fails to insert during a required scram.

Even with this postulated single failure, enough reactivity control remains to reach and maintain hot shutdown. Also, improved TS 3.1.3 RA A.2 requires a notch test on each withdrawn control rod to ensure no more control rods are stuck. With these changes, more time to verify SDM is needed to do the analysis or test.

2.

Existing TS 3/4.1.2 had the following less restrictive change:

a.

Improved TS 3.1.2 RA A.1 increases from 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> the existing TS 3.1.2 Action a. time allowed to explain and restore the.

core reactivity difference to within limits (existing TS: perform an analysis to determine and explain the cause of the reactivity difference). A reactivity anomaly normally indicates incorrect analysis inputs or assumptions of fuel reactivity used in the analysis. Contacting an offsite fuel analysis department and the fuel vendor and obtaining the necessary input may require a time period much longer than one shift. Since SDM is demonstrated by a test before reaching the conditions for this surveillance, the safety.

impact of the extended time for evaluation is negligible.

's i

b j

28 j

b.

The licensee proposes 'to replace " core alterations" in existing TS SR 4.1.2.a with " fuel movement within the reactor pressure vessel or j

l control rod replacement." The intent of this surveillance.is to 4

verify the core reactivity after in-vessel operations which could have significantly altered the core reactivity. Certain Core i

Alterations have a known effect which is reversible and, in fact, are activities consistent with those assumed to occur during routine 3

operations. Normal control rod movement is such an activity.

j Improved SR 3.1.2.1 provides a specific list.of those Core

-l s

Alterations which would constitute a core reactivity change not 1

expected to occur during normal operations, specifically excluding normal control rod movement.

1 J

j 3.

Existing TS 3/4.1.3.1 had.the following less restrictive changes:

I 2

a.

Improved TS-3.1.'s Condition D. provides the requirements and actions j

for the local distribution of inoperable control rods.

It addresses i

three changes-i (1) A Note modifies Condition D. by excluding its applicability i

above 10% thermal power. To preserve scram reactivity, a stuck rod must be separated from other withdrawn inoperable control i

rods that may also not scram. This is addressed in improved TS i

LCO 3.1.4..The improved TS imposes separation requirements when j

below 10% thermal power. The Control Rod Drop Accident (CRDA) in the Safety Analysis Report presents the concerns for control

{

rod reactivity worth when operating the plant in that power j

region. Above 10% thermal power, a CRDA resulting in fuel i

damage is not a credible transient and LCO 3.1.4 provides appropriate limitations on control rod separation.

i (2) Condition D. does not require action for inoperable control rods whose positions are in conformance with BPWS constraints, even 1

l if the inoperable control rods are within two cells of each l

other. Below 10% thermal power, the appropriate core reactivity and power distribution limits are controlled by maintaining control rod positions within the limits of BPWS and maintaining i

scram times within the limits of existing TS 3.1.3.2 and

[

improved TS 3.1.4.

Therefore, the limit on the local distribution of inoperable control rods that comply with BPWS is either overly restrictive or adequately controlled by the distribution restrictions for slow control rods.

1 (3) Finally, RA D.1 or D.2 allows 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to correct the situation l

before commencing a required shutdown, while existing TS 3.1.3.1 Action a.1 allows one hour. This increase recognizes the actual i

t operational steps involved on discovery of an inoperable control i

rod. The low probability of a CRDA during this time extension and the desire not to impose excessive time constraints on operator actions, which could lead to hasty corrective actions, j

make this extension to the action acceptable.

{-

l

i; e i

29 b.

Disarming a control rod involves actions by other than control room operating personnel after getting inside Primary Containment.

Existing TS 3.1.3.1 Action a.1.b) requires that all these actions be completed and the control room personnel confirm completion within one hour.

Improved TS 3.1.3 RA A.1 extends this time to two hours, recognizing the potential for excessive haste required to complete i

this task. The two hour time is not a significant safety concern, as l

the control rod is already in an acceptable position in accordance-with other improved TS ras, and the action to disarm is only for-4 i

preventing the potential for damage to the Control Rod Drive (CRD) mechanism.

4 i

c.

Improved TS 3.1.3 removes the upper limit on restoration time for one stuck control rod. By deleting existing TS 3.1.3.1 Action a.4 1

continued operation with a stuck control rod may be allowed. With a.

single withdrawn control rod stuck, the remaining operable control i

5 rods are capable of providing the required scram and shutdown reactivity. The assumptions utilized in establishing the proposed scram time limits account for a single stuck control rod in addition i

to an assumed single failure during a transient. 'SDM is required to 2

be met, accounting for the loss of negative reactivity due to the stuck control rod. Since operation continues to be within the bounds i

of analyzed events, the. improved TS require that.all limits be met t

plus prompt action be taken to confirm no additional stuck control rods exist, and therefore the improved TS allow continued operation l

and mode changes in accordance with SR 3.0.4.

l d.

Improved TS 3.1.3 Actions C.1 and C.2 require an inoperable non-stuck control rod to be inserted fully and disarmed.

It extends the time i

allowed to complete the insertion to three hours for all cases.

Existing TS LCO 3.1.3.4 Action a.2, for an uncoupled control rod, i

allows two hours before entering existing TS 3.1.3.1 Action b.1 that i

then gives one hour to insert and disarm the control rod (total of i

three hours to insert and disarm).

Improved TS 3.1.3 Condition C i

addresses uncoupled control rods and other non-stuck inoperable control rods. The existing TS 3-hour allowance, before requiring an i

i inoperable (uncoupled) control rod to be inserted, is the same found 2

j in the improved TS RA C.1 for control rod insertion. For consistency, the improved TS uses this three hour limit for all other i

instances of inoperable control rods. Given that these instances do i

i not represent a loss of SDM and are limited to a total of no more l

than eight inoperable control rods (improved TS Condition E), the i

l extended time does not represent a significant safety concern.

I Disarming a control rod takes actions by other than control room i

persons after getting inside primary containment. The existing TS requires these actions be completed and the control room persons confirm completion within the I hour allowed to insert the control rod. The improved TS extends this time to four hours, one~ hour beyond that allowed to insert, to reduce the potential for excessive 4

haste to do this. The four hour time is not a significant safety concern, as the control rod is already in its required position by 4

i i

30

]

i other improved TS ras and the action to disarm is_ only for preventing the potential for future incorrect operation.

1 1

e.

The improved TS move the SDV vent and drain valve requirements to 1

LCO 3.1.8.

The primary safety functions of these valves are to i

maintain the scram discharge volume with enough capacity to accept discharge water following a scram signal, and to: isolate the SDV during a scram to contain the reactor coolant discharge. The isolation function can still be satisfied if at least one valve is' 1

operable 'in each line or the line is isolated. Thus, the improved TS l

modifies the existing TS to:

(1) Allow seven days to restore inoperable SDV vent or drain valves provided at least one valve in each line.is operable or the line is isolated.

(2) Require isolation of the line in eight hours with _both valves 4

inoperable, concurrent with the seven day limit to restore both valves to operable status.

l i

(3) Recognizing that the SDV vent and drain valves are normally open to prevent accumulation'of water in the SDV from leakage, add a i

Note to RA B.1 to allow periodic opening of the isolated line l

for draining and venting of the SDV and avoid automatic reactor i

j scrans on high level in the SDV.

j l

These increased allowances and the option to open a SDV line j

isolated to comply with RA B.1 do not substantially increase the i

risk of a scram with an additional failure that could allow the

]

l SDV to remain open, nor substantially increase the risk of the l

SDV failing to accept the control rod drive water displaced j

during a scram.

i

[

f.

The existing TS 3.1.3.1.a.2 requirement to demonstrate shutdown margin with a control rod stuck is proposed as improved TS 3.1.3 RA j

A.3.

With a single control rod stuck in a withdrawn position, the i

i remaining operable control rods 'are capable of providing the required scram and shutdown reactivity.

Failure to reach cold shutdown is-only likely if an additional control rod adjacent to the stuck control rod also fails to insert during a required scram.

Even with the postulated additional single failure of an adjacent control rod l

to insert, sufficient reactivity control remains to reach and maintain hot shutdown conditions. Required Action A.2 of improved TS 3.1.3 performs a notch test on each remaining withdrawn control rod to ensure that no additional control rods are stuck. Given these

}

considerations, the time allowed to demonstrate shutdown margin has been extended from 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to allow a reasonable time to 4

perform the analysis or test.

g.

The existing TS SR 4.1.3.1.4.a requirement that SDV vent and drain valve closure and opening following a scram performance be from a normal control rod configuration of less than or equal to 50% rod l

i

_...-__.,_l

..,._....m

4 6

0 31~

i density, has-been deleted. The operability of the valves can be satisfactorily demonstrated during a scram from shutdown conditions.

j Reactor pressure and CRD discharge flow conditions do not influence

)

the SDV vent and drain valve closure rates since the SDV is of sufficient volume and initially vented such that peak pressure prior to SDV isolation will not be substantial. At less than or equal to 50% control rod density, back pressure would become significant j

following a test, but would not be significant during testing at shutdown conditions. The ability of the valves to open against.this back pressure is demonstrated after each reactor scram during 4

l operation. The lower coolant temperatures expected during testing at -

i shutdown conditions will also have a negligible impact on the l

performance of the surveillance. Therefore, since the surveillance can be satisfactorily performed at shutdown conditions, the 50%

control rod density requirement has been deleted. Additionally, 4

deleting this requirement eliminates the need for the existing TS SR j

4.0.4 not applicable statement.

l h.

After discovery of a stuck rod, all withdrawn control rods are required to be notch tested by improved TS 3.1.3 RA A.2.

This l

provides adequate assurance that the cause of the stuck rod is not a rooted in common cause failure of multiple control. rods. Continued testing of control rods at the normal 7 day improved TS SR 3.1.3.1 frequency 'is sufficient to ensure continued operability of. the remaining control-rods and therefore the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> frequency is only required once and not repeated at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as required by existing TS SR 4.1.3.1.2.b.

4.

Existing TS 3/4.1.3.2 had the following less restrictive changes:

a.

The Design Basis Accident and transient analyses assume all of the control rods scram at a specified insertion rate. The resulting l

negative reactivity forms the basis for determining plant thermal limits (e.g., the maximum critical. power ratio). Other distributions L

of scram times (i.e., several control rods scramming slower than the i

average time, and several control rods scramming faster than the j

average time) can' also provide sufficient scram reactivity within the j

assumptions of the analyses.

/

To account for single failure and to allow a certain number of slow i

scramming control rods, the scram times in improved TS Table 3.1.4-1 I

are faster than those assumed in the design basis analysis. These

]

scram times have sufficient margin to those assumed in the analyses to allow up to 7.5% of the control rods to have scram times that 1

exceed the specified limits. The improved TS assume they do not=

1 scram but insert in the long term, with insertion times in minutes or j

hours versus seconds.

It classifies control rods with scram times exceeding the LCO limits as slow. Also, the improved TS limits have 4

margin to account for a single stuck control rod allowed by improved TS 3.1.3 Condition A, and an additional control rod failing to scram per'the single failure criterion (total of two stuck control rods).

4

=

j.

32 Improved TS 3.1.4 changes the existing TS actions, according to the above discussion, as follows:

(1) The improved TS relaxes. the Action a.1 maximum limits on slow

)

control rods to seven seconds for notch position 13, with no intermediate limits for other notch positions.

It states this limit in Table 3.1.4-1 Note 2 and in SR 3.1.3.4.

It assumes these slow control rods do not scram.

It bases this seven=

second limit on historical use of this value, which bounds

{

assumptions for long term reactivity control.

1 l

(2) The improved TS LCO limits assume the existing TS Action a.2 l

maximum average scram times..This change is more restrictive in that.the existing TS maximum average limits are faster than the i

existing TS LCO limits.

(3) The improved TS increases the existing TS Actions a.3 and c.4 l

numbers of control rods exceeding the LCO limits (fast and slow -

control rods, respectively). The overall core scram reactivity remains consistent with that assumed in analyses. The allowed 4

number of slow control rods in the improved TS, 7.5% of the total number of control rods,. is higher than the 4% allowed in j

the existing TS because withdrawn inoperable control rods are no j

longer allowed.

}

(4) The improved TS requires the existing TS Actions b. and c.2, i

excessively slow control rods (i.e., with scram times in excess j

of seven seconds for notch position 13) be declared inoperable.

i The improved TS actions for inoperable control rods require these excessively slow control rods be fully inserted and 4

disarmed, while the existing TS actions' allow these control rods to remain withdrawn. While the criterion for excessively slow l

control rods is relaxed from values in existing TS Action a.1 to seven saconds, the accident analysis assumptions are still met.

j b.

The improved 15 deletes the existing TS Action b.2 requirement for increasing scr0m time surveillance testing when three or more control rods are slow. During normal power operating conditions scram j

testing is a significant perturbation to steady state operation which j

requires significant power reductions, abnormal control rod patterns, j

and abnormal control rod drive hydraulic system configurations.

Requiring more frequent scram time surveillance tests is therefore not desirable. Because the improved TS require frequent testing of j

control rod insertion capability (SR 3.1.3.2 and SR 3.1.3.3) and j

accumulator operability (SR 3.1.5.1), and the operating history demonstrates a high degree of reliability, the more frequent scram time testing is not deemed necessary to ensure safe plant operation.

i c.

The improved TS deletes the existing TS Action c.3.a) additional l

scram insertion testing to determine if an adjacent slow drive i

exists. Since scram insertion time data exist from previous testing i

for control rods surrounding a slow control rod, no more testing L

4 i

~

O o

33 should be required if the results of the existing TS Action c.1 (improved TS SR 3.1.4.3 and_ Bases) statistical-sampling of scram times show the criteria are met.

The improved TS limits on the number and distribution of slow rods are stringent enough to ensure that severe degradation of scram performance does not exist. Because the improved TS requires frequent testing of control rod insertion capability (SR 3.1.3.2 and SR 3.1.3.3) and. accumulator operability (SR 3.1.5.1) and the operating history demonstrates a high degree of reliability, the more frequent scram time testing is not deemed necessary to ensure safe plant operation.

d.

The improved TS modifies the existing TS 4.1.3.2.1.a. to require scram insertion time testing only after fuel movement within the reactor pressure vessel. This is equivalent to the existing TS definition.of core alterations, minus normal control rod movement.

Normal control rod movement with the vessel head detensioned or

. removed, however, is not any different than normal control rod movement _during power operations.

It does not affect scram speed.

Therefore, no increased frequency of scram time test performance is necessary simply due to normal control rod movement while the vessel head is detensioned.

5.

Existing TS 3/4.1.3.3 had the following less restrictive. changes:

a.

An inoperable control rod accumulator affects the associated control rod scram time. However, at sufficiently high reactor pressure, the accumulators only provide a portion of the scram force. With this reactor pressure, the control rod will scram even without the associated accumulator, although probably not within the required scram times. Therefore, im> roved TS 3.1.5 allows the option of declaring a control rod witt an inoperable accumulator slow when reactor pressure is sufficient.

Since the existing TS action to declare the control rod inoperable would allow the control rod to remain withdrawn and not disarmed, the improved TS action to. declare the control rod slow is essentially equivalent.

Improved TS LCOs 3.1.3 and 3.1.4 appropriately apply limits and allowances for the numbers and distribution of inoperable and slow control rods to control rods with inoperable accumulators whether declared inoperable or slow. Notes to improved TS 3.1.5 ras A.1 and B.2.1 restrict the option for declaring the control rod with an inoperable accumulator slow to control rods that were not previously known to be. slow. This restriction limits the flexibility to control rods not otherwise known to have an impaired scram capability.

Also, improved TS 3.1.5 ras B.2.1 and B.2.2 extend to one hour the-time for declaration of slow or inoperable control rod (and the implied concurrent allowed control rod restoration time). This provides a reasonable time to attempt investigation and restoration of the, inoperable accumulator. Also, the improved TS makes_the actions more user friendly by presenting first the more critical action (i.e., with a shorter allowed Completion Time).

Improved TS RA B.1 addresses the situation where additional accumulators may be -

l l

i

l l

i 4 -

34 i

rapidly becoming inoperable due to loss of charging pressure. Once j

verification of_ adequate charging pressure is made, and considering i

that reactor pressure is adequate to assure the scram function of the control rods with inoperable accumulators, the proposed one hour extension is not significant.

L i

l.

b.

Existing TS Action a.2.a) for inoperable scram accumulators applies 1

)

to all reactor pressure situations, whether at normal operating j

pressure or zero pressure. These two extremes represent significant differences in whether a control rod with an inoperable accumulator i

i will scram or not. The improved TS ras acknowledge this difference-1

(

and present actions more appropriate to the actual plant conditions (in one instance, proposing more restrictive actions).

4 Existing TS Action a.2.a) is intended to identify the situation where i

additional scram accumulators (eventually all accumulators) would be i

expected to become inoperable.

Identification of this sort of common cause is significant in ensuring continued plant safety.

In the i

event reactor pressure is too low, where the control rod with an i

inoperable accumulator may not scram, it is imperative that immediate j

action be taken if the charging pressure to all accumulators is lost.

Improved TS RA C.1 maintains this action essentially consistent with 4

the existing TS action.

However, in the event reactor pressure is sufficiently high (where j

the control rod will scram even without the associated accumulator),

1 20 minutes is proposed to ensure that control rod accumulator charging water pressure is adequate to support maintaining the remaining accumulators operable. This 20 minutesiallows an i

appropriate time to attempt restoration of charging pressure if it should be lost._ This action is deemed more appropriate than the option to initiate an immediate reactor scram. The most likely cause of the loss of charging pressure is a trip of the operating CRD pump.

l Restart of this pump or of the spare CRD pump would restore charging pressure and avoid the probability of a plant transient caused by the 4

immediate scram initiated while withdrawn control rods with i

inoperable accumulators are known to exist, and the system necessary I

for manual control rod insertion is not available. Since control rod 4

i scram capability remains viable solely from the operating reactor

}

pressure, and the most likely result of the 20 minute allowance is J

expected to be restoration of charging pressure (at which time inoperable control rods could be manually inserted and disarmed, i

operation returned to normal, and a scram transient avoided), this less restrictive change is deemed acceptable.

6.

Existing TS 3/4.1.3.4 had the following less restrictive changes:

a.

If the applicable rod pattern control system does not allow an uncoupled control rod to be inserted to accomplish recoupling, the existing TS requires the control rod to be inserted. This may also require byparsing of the applicable rod pattern control system and operation with an out-of-sequence control rod. Therefore, in the 4

s i

l C

,,.-_,m

=

i i

5 35 improved TS coupling attempts are allowed regardless of the condition of the rod pattern control system because of the short time allowed.

If coupling is not established in three hours, improved TS 3.1.3 ras C.1 and C.2 require the control rod to be fully inserted then 5

disarmed within the next hour. Also, because of the limited time j

allowed to recouple, the improved TS do not restrict the number of.

j attempts, although plant procedures that consider the potential for equipment damage during successive recoupling attempts may restrict the number of attempts to recouple a control rod.

4 i

b.

Existing TS 3.1.3.4 Action b. coupling requirements during refueling i

are not necessary since only one control rod can be withdrawn from i

core cells containing fuel assemblies. The probability and S

consequences of a single control rod dropping from its fully inserted -

position to the withdrawn position of. the control rod drive are j

negligible (i.e., reactor will remain subcritical). However, these requirements are retained for the improved TS 3.10.8, Special

]_

Operation of shutdown margin testing in Mode 5.

j

.7.

Existing TS 3/4.1.3.5 had the following less restrictive changes:

3 a.

The existing TS does not clearly present the time allowed for inserting and disarming a control rod after declaring a control rod inoperable. The requirement follows an action to " declare the control rod inoperable" and has been interpreted to give direction as.

to the proper option for inoperable control rod actions.

Existing LCO 3.13.1, Action b.1 allows one hour from the time the control rod l

1s declared inoperable, until it is required to be inserted and disamed. Therefore, the intent of the existing TS requirement is deemed to allow a total of two hours; one hour in existing TS 3.i.3.1 Action a, plus the one hour in LC0 3.1.3.5.

Improved TS 3.1.3 Condition C. addresses control rods whose position is unknown, as well as other non-stuck inoperable control rods. The l

improved TS 3.1.3 RA C.1 extends to three hours the existing 1S two hour allowance, before requiring an inoperable (position unknown) rad i

to be inserted. For consistency of presentation,'the improved TS uses this three hour limit for all other instances of inoperable i

control rods. These other instances (excessive scram speed, and i

certain combinations of conditions with a low pressure on a control rod scram accumulator) also warrant a minimal time to attempt j

restoration before inserting and disarming. Since these and unknown control rod position do not represent loss of SDM and improved TS 3.1.3 Condition E limits the total inoperable control rods to no more 1

l than eight, the extended time is not a significant safety concern.

Disarming a control rod takes actions by other than control room persons after getting inside' primary containment. The existing TS requires that these actions be completed and the control room persons confirm completion within the one hour allowed to insert the control rod. The improved TS extends this time to four hours, one hour beyond that allowed to insert, to reduce the potential for

/

}

..~ i

j '.

1 i

l 36 i

excessive haste to do this. The four hour time is not a significant i

safety concern, as the control rod-is already in.its required

! ~

position by other improved TS ras and'the action to disarm is only for preventing the potential for future incorrect operation.

b.

Improved TS SR 3.1.3.1 deletes existing TS 4.1.3.5 d. that requires a i

12-hour channel check when the alternate control rod position j

i indicator is operable. The Rod Pattern Control System (RPCS).

i receives control rod position inputs from two independent channels.

i The RPCS continuously performs a self test to confirm agreement j

between the two channels (i.e., a channel check).

Should either channel's position input become unknown, or should the two channels disagree on position, or the channel check otherwise fail, or the 1

i.

test circuit fail to perform any. self test, the system initiates a l

control rod block and alarms the malfunction.

Deleting existing TS SR 4.1.3.5 d. will not have an impact on the reliability of the system, on the ability of the operator to detect a control rod whose l

position is unknown, or on safe operation of the plant.

l 8.

Existing TS 3/4.1.4.2 had the following less restrictive changes:

i 3

The designed RPCS limits control rod. patterns and movement to.within the bounds of analyzed conditions.

If the system is inoperable, then the o

actions presented in' existing TS Action a. appropriately apply. These j

actions will automatically occur by the fail-safe nature of. the RPCS in the majority of conceivable RPCS inoperabilities.-

i However, the requirements of existing TS Action b. contain conditions j

that may not be directly related to the RPCS function, and therefore, do not automatically warrant the same actions. When these conditions result i

in operation beyond that allowed by the design of the RPCS, the RPCS will i

either preclude their occurrence (by blocking control rod movement) or j

the system would be considered inoperable and the appropriate actions applied consistent with existing TS Actions a.1 and a.2.

When the i

condition does not result in a situation that the RPCS is intended to

{

j preclude, the improved TS has other more appropriate actions. Other i

j existing TS Chapter 3.1 LCOs/ Actions and improved TS Chapter 3.1 1

LCOs/ Actions address the following conditions not controlled by the RPCS, as discussed below:

l

+

l (1) Existing TS Action b.1, addresses determination of SDM with a stuck control rod.

Existing TS 3.1.1 and 3.1.3.1 Actions also address this condition. The RPCS does not support maintaining SDM when a control J

rod is stuck, and should not require actions to suspend control rod movement until SDM has been confirmed. Other specifications provide appropriate action for this condition.

Improved TS 3.1.3 specifies

}

separate actions, allowing 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for re-determining the SDM. The assumptions in the safety analyses justify this time, accounting for j

the effects of one stuck control rod. Allowing the control rod to be j

bypassed during this 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period does not introduce any 4

i

)-

... - -,.. ~.... - -

i i

2 37 i

significant impact on safety for continued operation of the plant j

while determination of the SDM is being made.

I (2) Existing TS Action b.2.a) restriction, positioning and disarming of j

an inoperable rod, appears to be incorrectly worded. The provision allows bypassing and continued operations only after an inoperable control rod is inserted.

In actual practice, to complete the insertion may require bypassing the control rod.

In either case, 1

improved.TS 3.1.3'provides the necessary actions for inserting and-j disarming an inoperable control rod. The improved TS change here allows the bypassing of inoperable control rods, at:any time, and allows other specifications to control the insertion and disarming of l

that control rod. Since the more safety significant action would be ~

to address the inoperable control rod, if bypassing it prior to i

insertion would facilitate that action, it is not a significant safety' issue to allow that bypassing.

{

(3) Existing TS Action b.2.b) which discusses inoperable control rod 1

i separation criteria, is not necessary in all cases.

Improved TS 3.1.3 Condition D. adequately controls' the separation requirement J

below 10% thermal power.

Improved TS 3.I.3 Condition C. requires

{

that an un-stuck inoperable control rod be inserted. Two such rods may be adjacent if their respective BPWS groups are allowed to be e

inserted.

If not in compliance with BPWS, improved TS 3.1.3 j

Condition E. would require a shutdown.

Existing TS Actions a.l. and b.2.b) require a shutdown by scram when below the low power setpoint j

in all cases of inoperable control rods not separated. A scram-initiates an unnecessary challenge to'the equipment that mitigates f

I such a transient.

Improved TS 3.1.3 Condition D. allows four hours j

to attempt to return to compliance with BPWS or return the rod (s) to operable status. This time is based on the Condition C. requirement l

to fully insert an inoperable rod. This is based on the high probability of success and the low probability of a CRDA during this time frame. A controlled shutdown, if needed, prevents an l'

unnecessary challenge to the equipment that mitigates a transient caused by a scram.

l (4) The improved TS delete existing TS 3.1.4.2 Actions b.t.c) and b.3.b) that require no more than three inoperable control rods in any RPCS group. Plant specific analyses have been performed to eliminate the i

restriction on the number of inoperable control rods in a BPWS group.

1 (The BPWS is the analytical, procedural approach to control rod j

withdrawal, in compliance with the RPCS instrumentation required i

operable by improved TS 3.3.2.1, where the improved TS relocated most 1

of existing TS 3.1.4.2.)

A BPWS group can allow up to eight inoperable rods if the required separation is maintained.

Improved TS 3.1.3 Condition D. and 3.1.6 Conditions A. and B. contain these requirements. The separation requirements force the inoperable i

i control rods to be evenly distributed, which minimizes the i

consequences of the out-of-sequence rods. The bounding analysis l

performed to support this change uses the previously approved j

methodology for a Control Rod Drop Accident analysis. As identified i

+

_.,c.

o I[

38 L

l in existing TS 6.9.1.11, this methodology is currently approved for use in establishing cycle-specific operating limits.

s (5) Improved TS 3.1.6 contains a specific requirement for control rods to be in compliance with the BPWS during operation:at low power. This improved TS Condition A also contains an allowance for a limited i

number of out-of-sequence operable control rods. ' The' condition allows up to eight of these control rods (as opposed to inoperable out-of-sequence control rods) to be returned to their correct t

position within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />., This allowance for correction recognizes the occurrence of such events as double-notch rod withdrawals, and -

minor misalignment of rod pattern.during CRD hydraulic transients (control rod drift due to excessive cooling water pressure) or during a

a plant shutdown. These events can introduce out-of-sequence control i

rod patterns, which the RPCS was unable to preclude, even though the RPCS was' functioning as desigt4d.

l Existing TS 3.1.4.2 Actions b.3 places the following restrictions on the restoration of out-of-sequence control rods:-1) out-of-sequence j

and/or inoperable control rods must be separated; 2) no more than i

three out-of-sequence and/or inoperable control rods in any RPCS l

group; and 3) a limit of 8 control rods may be bypassed and/or.

out-of-sequence at a time.

If these restrictions cannot.be satisfied i

at low power, the existing TS Actions a.1 require a reactor scram as the only pemissible control rod movement. Since these out-of-sequence control rods are operable, and.the RPCS is operable (if not, improved TS 3.3.2.1 RA B.1 would require a scram as the only permissible control rod movement), the control rod pattern could be restored to compliance with BPWS in a brief period of time.

Each control rod moved during the correction process would be required to 4

be bypassed, which in turn would require a second verification of the

]

proper movement of the control rod (see improved TS SR 3.3.2.1.9).

I The improved TS 3.1.6 RA A.18-hour time allowed for correction is small, with each step bringing the pattern closer to compliance with j

BPWS. The probability of a Control Rod Drop Accident is remote during this brief period. Therefore, this time allowance is deemed to not present a significant impact on safe operation, and would 4

j preclude the plant transient introduced by a required reactor scram.

9.

Existing TS 3/4.1.5 had the following less restrictive changes:

1 1

(1) The existing TS applicability includes control rod withdrawal in Mode 5.

The standby liquid control (SLC) system provides the capability i

of bringing the reactor, at any time in a fuel cycle, from full power i

and minimum control rod inventory to a subcritical condition with the reactor in the most reactive xenon free state without taking credit for normal control rod movement. The SLC system is used in the event not enough control rods can be inserted to accomplish shut down and i

cool down in the normal manner.

In Mode 5, only a single control rod can be withdrawn from a core cell containing fuel assemblies.

Demonstration of adequate SDM during control rod withdrawal in Mode 5 i

l j

4 39 prevents criticality from occurring during these conditions. Based on the above discussion, the need to avert criticality events during i

single control rod withdrawal is adequately controlled by the SDM demonstration; therefore, the existing TS requirements for the SLC system backup reactivity control capability are not required for the

{

improved TS 3.1.7.

(2) Existing TS 4.1.5.a.3 includes surveillances that test heat tracing circuitry and pump suction piping temperature every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. These tests are adequately covered by daily verification that the pump suction piping is within the temperature limits as required by improved TS SR 3.1.7.3.

(3) An additional allowance for a valve in the flow path to be capable of being aligned to the correct position is provided in improved TS SR 4

3.1.7.6 to satisfy existing TS SR 4.1.5.b.4 flow path alignment. This allowance may be generically added for any system which is manually initiated. Operktor action will provide system initiation consistent

)

with the safety analysis.

(4) Improved TS SR 3.1.7.2 daily surveillance testing of the solution temperature provides an adequate check on the capability of the storage tank heaters to maintain solution temperature and therefore i

the 18 month test of the tank heaters in existing TS SR 4.1.5.d 4 is i

i deleted.

i 1

The above less restrictive requirements have been reviewed by the staff and have been found to be acceptable, because they do not present a significant j

l safety question in the operation of the plant. The TS requirements that i

remain are consistent with current licensing practices, operating experience and plant accident and transient analyses, and provide reasonable assurance that the public health and safety will be protected.

1 3.2 Power Distribution Limits A.

Significant Administrative Changes.

In accordance with the guidance in the Final Policy Statement, the licensee has proposed administrative changes to the existing TS to bring them into conformance with the improved TS. The licensee has elected not to include improved TS 3.2.4, " Average Power Range Monitor (APRM) Gain and Setpoints (Optional). " Deletion of the APRM Gain and Setpoints specification was i

i previously submitted by the licensee based on analysis for the maximum extended operating domain and approved as Amendment 16 to the GGNS Operating License. Therefore, the staff finds that the improved TS conform to the current licensing basis and the deletion of improved TS 3.2.4 is acceptable.

i l

40 l

B.

Relocated Requirements 1

In accordance with the guidance in the Final Policy Statement, the licensee has proposed to relocate all or portions of the following existing TS to other licensee-controlled documents:

i Existina TS Iltle 3/4.2.1 Average Planar Linear Heat Generation Rate

_j j

3/4.2.3 Minimum Critical Power Ratio i

1 3/4.2.4 Linear Heat Generation Rate i

i The more significant changes resulting from relocated items are as follows:

i i

The licensee has proposed to relocate to the TS Bases or reorganize all or portions of the existing TS Section 3.2 action statements which deal with the requirement to initiate corrective action within 15 minutes.

The discussion in the Bases will describe the prompt action to be taken to restore the parameter to within limits.

Corrective action in 15 minutes may not always be the conservative method to assure safety. The two hour completion time allows the operator-sufficient time to evaluate j

and complete appropriate actions.

i l

The above relocated requirements relating to power distribution limits, fuel j

rod minimum critical power ratio limits and fuel rod linear heat generation -

rates are not required to be in the TS under 10 CFR 50.36, and are not t

required to obviate the possibility of an abnormal situation or event giving j

rise to an immediate threat to the public health and safety.

Further, they do j

not fall within any of the four criteria set forth in the Commission's Final Policy Statement, discussed in the Introduction above.

In addition, the staff finds that sufficient regulatory controls exist under 10 CFR 50.59.

Accordingly, the staff has concluded that these requirements may be relocated from the TS to the licensee's UFSAR.

l C.

Less Restrictive Requirements j

The licensee, in electing to implement the NUREG-1434 Section 3.2 j

Specifications, has proposed certain less restrictive conditions than are required by the existing TS. The more significant conditions are the j

following:

i 1.

Existing TS 4.2.1.b, 4.2.3.b, and 4.2.4.b require that a surveillance be l

performed after every 15% power change or at the end of any single power 1

increase greater than 15%. The improved TS eliminates confusion as to i

how often to perform the surveillance and allows a single verification during initial startup considering the large inherent margin to operating

[

limits at low power levels. Verifying the required parameter within 12 1

hours of reaching or exceeding 25% rated thermal power will generally require that the surveillance be performed after completion of a 15%

i power increase, but would also reduce the number of times the surveillance must be conducted during a startup if it is currently conducted after every 15% power change. Following the initial verification, the improved TS requires the surveillance be performed d

I i

-~

i i

41 i

every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to identify any trends in these parameters that may lead i

to long ters noncompliance.

I 2.

The improved TS delete existing TS 4.2.1.c, 4.2.3.c and 4.2.4.c which a

deal with verifications while operating with a limiting control rod

^

pattern. Since a limiting control rod pattern is currently defined as a control rod pattern that results in operating on a power distribution limit such as Average Planar Linear Heat Generation Rate or Minimum Critical Power Ratio, the condition is extremely unlikely, and the surveillance would almost never be required. Also, since it would not be 3

evident that a limiting control rod pattern has been achieved until the surveillance is performed, the initial surveillance is superfluous.

l The above less restrictive requirements have been reviewed by.the staff and have been found acceptable because they do not present a significant safety i

question in the operation of the plant since the TS requirements that remain l'

are consistent with current licensing practices, operating experience, and plint accident and transient analyses.

3.3 Instrumentation j

A.

Significant Administrative Changes l

In accordance with the guidance in the Final Policy Statement, the licensee has proposed administrative changes to-the existing TS to bring them into confermance with the improved TS. These changes are as follows:

1.

Existing TS 3/4.3.1 footnotes that provide remedial action requirements -

l for channel and trip system inoperability have been incorporated into the

-l improved TS actions consistent with the guidance of NUREG-1434. The existing TS footnotes provide instructions to the operators such as an i

inoperable channel need not be placed in the tripped. condition where this i

would cause the trip function to occur.

It is the presentation preference of the guidance in NUREG-1434 to explicitly state these instructions. The improved TS incorporates other significant operator l

instructions by including appropriate conditions and required actions based on combinations of single channel, multiple channel and trip system inoperabilities to address loss of redundancy protection and loss of safety system functional capability. The presentation of these i

requirements is an administrative change with the requirements i

effectively retained in the improved TS, and is therefore acceptable.

i 2.

Existing TS 3/4.3.3.1 requires all of the channels from each of the 8 ~

4 l

MSIVs that supply closure signals to RPS trip systems to be operable.

All channels are required operable to ensure that a scram will follow the worst single failure. Therefore, improved TS 3.3.3.1 specifies eight as the minimum number of channels. This change in nomenclature is i

administrative because it does not affect'the design.

i 3.

In a letter of May 20, 1993 (GNRO-93/00049), the licensee proposed changes that modify the hot shutdown and the refueling mode

4 42 l

applicability requirements for neutron flux high and neutron-flux inoperative functions of the RPS. This change was approved by the staff

,[

in Amendment Number 109, dated December 13, 1993.

4

}

4.

Existing TS 3/4.3.2 includes a footnote that provides'a description of when the condenser vacuum main steam isolation valve closure can be L

bypassed and when the bypass must be removed. The improved TS footnote (a) of. Table 3.3.6.1-1 describes the intended applicable conditions more succinctly. Rather than a requirement for operability with a note that allows bypass, the applicable condition limits the conditions where the i

function could not be bypassed. Since this results in a requirement for the capability to perform its safety function in the same applicable l

conditions, this change is considered administrative.

i 5.

The existing TS requirement to declare the affected system inoperable I

when a penetration flow path is isolated because less than the required 1

minimum number of channels are operable (Table 3.3.2-1) is essentially retained since improved TS LCO 3.0.6 combined with improved TS 3.3.6.1 i

actions specify an equivalent requirement.

In this case, the LC0 3.0.6

{

supported system _(e.g., LPCI) is made inoperable by isolating the penetration flow path not by the inoperable instrumentation. - Therefore.

l LCO 3.0.6 does not prevent the LPCI system actions from being taken.

Thus, equivalent directions in the instrumentation actions are provided l

to the operator that more precisely states the intent of the existing TS l

action.

i 6.

Improved TS 3.3.6.1 includes ras H.1 and H.2 requiring shutdown if the l

affected penetration flowpath(s) are not isolated within the allowed l

completion times of ras F and G.

This requirement is in agreement with existing TS 3/4.3.2. The improved TS LCO 3.0.3 which requires shutdown i

if LCO actions are not satisfied does not constitute a change in l

requirements, and is acceptable.

4 7.

Existing TS 3/4.3.2 Action 25 includes the term secondary containment i

integrity. The improved TS replaces the defined term secondary containment integrity with the essential elements of the definition. The change is editorial in that all the individual requirements are i

specifically addressed by the proposed required actions. Therefore, the

[

change is' purely a presentation preference consistent with the guidance provided in NUREG-1434.

i i

8.

All loss of power (LOP) instrumentation requirements (i.e., loss of voltage, degraded voltage, and the allowable time delays) specified in i

existing TS 3/4.3.3 are provided as separate items in improved TS 3.3.8.1.

The change is for presentation preference only and it is-

]

therefore acceptable since the requirements are unchanged due to this separation.

e 9.

The licensee proposes moving existing SR 4.3.3.3 requirements that specify time limits for HPCS response to initiation setpoints to improved i

i TS 3.5.1, ECCS-Operating and improved TS 3.5.2, ECCS-Shutdown.

ECCS l

response time operability requirements specify a time limit for the i

5 1

j*

i i

43 entire channel, from the time the monitored parameter exceeds its setpoint until the HPCS equipment is capable of performing its intended function. The SR includes testing of the-channel portion of the j

instrument on a staggered basis. This piece of the existing SR is relocated to the TS Bases, along with Bases discussion that associated actuation instrumentation testing is controlled by improved TS 3.3.5.1.

i The relocated requirements are a format change to the location of the HPCS system response time limits and are acceptable.

]

10.

In a letter of August 11, 1993 (GNRO-93/00088), the licensee proposed changes to control the requirements for the loose-part detection system of existing TS 3/4.3.7.10 in accordance with 10 CFR 50.59. This change l

was approved by the staff in Amendment Number 117, dated February 16, j

1995.

i I

11. Existing TS 3/4.3.4.2. LCO Applicability requirements were modified to i

include the phrase Rated Thermal Power (RTP) with any recirculation pump i

in fast speed. The End-of-Cycle (EOC) Recirculation Pump Trip (RPT)-

logic only trips the recirculation pump trip breaker for fast speed. The existing TS applicability is clarified in improved TS 3.3.4.1 to limit l

the requirements for this logic to those conditions during which its j

function is needed. Since this change represents a clarification of the original intent, this change is considered administrative and is therefore acceptable.

12.

In a letter of August 11, 1993 (GNRO-93/00097), the licensee proposed hanges that removed PAM functions from the PAM TS. This change was approved by the staff in Amendment Number 118, dated February 16, 1995.

2 As such this change is considered administrative in this submittal.

)

13.

Improved TS 3.3.1.2 includes Mode 5 Channel Calibration requirements that l

were not included in the existing TS 3.3.7.6.

The existing TS require l

Source Range Monitors (SRMs) be calibrated once every 18 months for 1

operability in Modes 2, 3, and.4. However, these are the same monitors that are used in Mode 5 for which the other (Modes 2, 3, and 4) Channel Calibrations are currently credited to provide assurance that the SRMs will properly perform their required function. Therefore, including the calibration surveillance for Mode 5 is an application of a currently l

required surveillance and, therefore, is an administrative change.

4 l

14. Existing TS 3/4.3.7.6 action requirements for Modes 3 and 4 were changed in the improved'TS from verify all insertable control rods to be fully inserted in the core to fully insert all insertable control rods. A single control rod may have been withdrawn under the provisions of the i

Special Operations LCO 3.10.3 and LCO 3.10.4, or some unanticipated event may have resulted in uninserted control rods. Therefore, rather than an action to verify all insertable control rods to be fully inserted, the improved TS required action is more-definitive; initiate action.to insert all insertable control rods. This wording provides the same intent in the event all insertable control rods are found to be inserted, but also-clarifies that any uninserted control rods are to be inserted.

4

+,.

,.m.

_.m.,,.

,._.,_..._s.,,,

,...e%n,-,.,,,,..-,,y,,.

.,...,.,..p,

l..

i 44 Since these requirements result in the same limits as the current-requirements, the changes are purely administrative and are therefore acceptable.

B.

Relocated Requirements In accordance with the guidance in the Final Policy Statement, the licensee j

has proposed to relocate or reorganize all or portions of the following.

j existing TS to other licensee-controlled documents:

i Existina TS IltJg l

l 3/4.3.1 Reactor Protection System Instrumentation 3/4.3.2 Isolation Actuation Instrumentation 3/4.3.3 Emergency Core Cooling System Actuation Instrumentation i

3/4.3.4.1 ATWS Recirculation Pump Trip Actuation i

i Instrumentation 3/4.3.4.2 End-of-Cycle Recirculation Pump Trip System i

Instrumentation 3/4.3.5 Reactor. Core Isolation Cooling System Actuation Instrumentation 3/4.3.6 Control Rod Block Instrumentation i

3/4.3.7.1 Radiation Monitoring Instrumentation i

1 3/4.3.7.2 Seismic Monitoring Instrumentation 3/4.3.7.3 Meteorological Monitoring Instrumentation i

3/4.3.7.4 Remote Shutdown System Instrumentation and Controls 3/4.3.7.5 Accident Monitoring Instrumentation 3/4.3.7.6 Source Range Monitors 3/4.3.7.7 Traversing In-Core Probe System 3/4.3.7.12 Main Condenser Offgas Treatment -Explosive Gas i

i Monitoring System-3/4.3.8 Plant Systems Actuation Instrumentation i

3/4.3.9 Turbine Overspeed Protection System l

The more significant changes resulting from relocated items are as follows:

i 1.

Existing TS LCO 3.3.2 includes Conditions, Actions, and Surveillance Requirements (SR) for Area Differential Temperature - High isolation i

actuation functions. The improved TS have relocated these functions to the UFSAR. The differential temperature instruments are designed to provide the capability to detect and initiate isolation of a 25 gpm-

)

equivalent steam leak. However, the differential temperature instruments constitute only one method of detecting steam leakage.

In addition to using area temperature monitors to detect reactor coolant system (RCS) i leakage, excessive leakage can also be detected by low reactor vessel water level, high process line flow, high differential flow, and various other plant specific monitoring methods. The differential temperature instruments are redundant to TS required area temperature monitors, level instruments and high flow instruments and do not contribute to the primary success paths to prevent or mitigate design basis accidents and 4

transients.

i

)

j.-

i 45 1

2.

Existing TS 3/4.3.3 includes a SR to perform a channel functional test (CFT) to verify that the automatic depressurization system (ADS) manual l

inhibit function is operable. This requirement was relocated to the UFSAR. The assurance that the ADS trip system is not rendered inoperable by the ADS inhibit function is tested in improved TS 3.3.5.1 by the logic system functional test. The ADS manual inhibit function provides the 4

capability to mitigate Anticipated Transient Without Scram (ATWS) event i

i by preventing ECCS initiation; however, the function is not required for operability of ECCS actuation instrumer.tation; i.e., initiate ECCS to-t preserve the integrity of the fuel cladding.

l 3.

The existing TS LCO 3.3.6 Conditions, Actions and SRs for the Average Power Range Monitors (APRMs), SRM, Intermediate Range Monitors.(IRMs),

scram discharge volume, and reactor coolant system recirculation flow control rod blocks are relocated to the UFSAR. During high power j

operation, the Rod Withdrawal Limiter (RWL) provides protection for control rod withdrawal error events. During low power operations,.

control rod blocks from the Rod Pattern Controller (RPC) enforce specific 4

i control rod sequences designed to mitigate the consequences of the CRDA.

4 During shutdown conditions, control rod blocks from the Reactor Mode i

Switch Shutdown Position ensure that all control rods remain inserted to i

prevent inadvertent criticalities. The APRM, SRM, IRM, scram discharge i

volume control and reactor recirculation flow rod blocks provide redundant capability to the TS required RPC, RWL and Reactor Mode Switch i.

functions for preventing control rod withdrawal errors at power-i transients.

4.

The existing TS 3/4.3.7.1 instrumentation functions for radiation l

monitors are relocated to the UFSAR. TS process parameters that indicate possible gross failure of fuel cladding, that a release may have i

originated from the primary containment or that a break occurred in the reactor coolant pressure boundary include: component cooling water, _

j standby service water system, plant service water, carbon bed vault, fuel i

handling area ventilation exhaust, fuel handling area pool sweep exhaust, fuel storage area, control room area, containment and drywell j

ventilation exhaust radiation detection functions.

I.

The detection of high radiation in the area surrounding the carbon bed vault or in the water stream of systems that are a boundary to the j

release of fission products is provided as an indication of local leakage. The plant safety analysis assumes other instrument functions a

i detect and isolate primary containment penetrations and penetrations I

which bypass secondary containment during a fuel handling accident.

}

Operability of instrumentation for ensuring release limits are met by initiating containment isolation are required by LCO 3.3.6.1 and LCO j

3.3.6.2.

i 1

The relocated fuel handling area exhaust, fuel handling storage area and containment and drywell ventilation exhaust radiation monitors provide a alarm that does not contribute to the automatic capability to isolate s

i containment ~ during a fuel handling accident, break in the reactor coolant j

pressure boundary or a condition of high exhaust radiation.

e t'

i 46 The. ability of the control room ventilation system to maintain the habitability of the main control room ensures that the radiation exposure j-of control room personnel, through the duration of the postulated accidents explicitly assumed in UFSAR Sections 6.4 and 15, does not 4

exceed the limits set by GDC 19 of 10.CFR Part 50, Appendix A.

The control room air intake radiation monitors automatically initiate action i

to route makeup air to the main control room through emergency filter units to minimize the consequences of radioactive material in the control i-i room. Operability of these monitors is required by LCO 3.3.7.1.

The 4

operability.of the relocated control room area radiation monitor provides an alarm that does not contribute to the capability to isolate the i

control room ventilation system from the source of high radiation.

i 5.

The existing TS 3/4.3.7.2 Conditions,-Actions, and SRs for seismic-i monitoring instrumentation are relocated to the UFSAR. -The seismic-j monitoring instrumentation provides monitoring capability by recording j

information regarding the severity of' an earthquake to permit comparison of the measured response to that used in the design basis.of the facility to determine if the plant can continue to be_ operated safely and to i

permit such timely action as may be appropriate pursuant to 10 CFR Part-i 100, Appendix A.

The requirements do not address the need for seismic-l monitoring instrumentation that would automatically shut down the plant j

when an earthquake occurs which exceeds a predetermined intensity.

t 6.

The existing TS 3/4.3.7.3 Conditions, Actions, and SRs for the i-meteorological monitoring instrumentation were relocated to'the UFSAR.

1 The meteorological monitoring instrumentation. is used to measure I

environmental parameters (wind direction, speed, and air temperature differences) which may affect distribution of fission products and gases following a design basis accident to be used in connection with the plans i

for coping with radiological emergencies, pursuant to 10 CFR 50.34(b),

and to provide a basis for estimating maximum potential annual radiation doses resulting from radioactive materials released in gaseous effluents, pursuant to 10 CFR 50.36a(a)(2).

7.

The existing TS 3/4.3.7.5 requirement to prepare and submit.a report to the Commission regarding the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to.

operable status are relocated to the improved TS Bases, i

i 8.

The existing TS 3/4.3.7.6 Action b. requires insertable rods to be inserted and the reactor mode switch to be locked in the shutdown l

position when one or more of the required SRM channels are inoperable in t

Modes 3 and 4.

The requirement to lock the reactor mode switch is i

relocated to the improved TS Bases. The-position of the reactor mode switch is defined by the Modes Definition Table (improved TS Table 4

1.1-1).

Movement of the reactor mode switch between defined positions results in changes in TS applicabilities. Therefore, any requirement to lock the mode switch in a specified position does not result in a change to TS applicabilities and can be controlled by plant procedures without physical locks to prevent violating procedural requirements. Mode switch l

b

f i

47 l

1 operability requirements are retained in the TS to ensure functions that i

depend on the mode switch position interlocks are also operable.

i 9.

The existing TS 3/4.3.7.7 Conditions, Actions, and SRs for the Traversing In-core Probe (TIP) has been relocated to the UFSAR. The TIP System is used to calibrate the Local Power Range Monitor (LPRM) detectors by-4 positioning the TIP axially and radially throughout the core. When not j

in use, TIP instruments are retracted into a storage position inside the drywell wall TIP penetrations during conditions for which the LPRM are required to be operable.

10. The existing TS 3/4.3.7.12 Conditions, Actions, and SRs have been j

relocated to the UFSAR. The Main Condenser Offgas Treatment System Explosive Gas Monitoring instrumentation is used to detect hydrogen in the main condenser offgas treatment system to ensure that hydrogen concentrations are maintained below the flammability limit. The offgas i

system, located in the turbine building, is designed to confine j

detonations without affecting safety-related equipment. The concentration of hydrogen in the offgas stream is not an initial 1

assumption of any Design Basis Accident (DBA) or transient analysis. The

]

relocation of the main condenser offgas treatment system, explosive gas i

monitoring system instrumentation is consistent with the presentation in NUREG-1434.

i;

11. The existing TS 3/4.3.8 feedwater pump / turbine trip on reactor vessel i

water level-high level 8 Conditions, Actions, and SRs were relocated to i

the UFSAR. Although the GGNS design has a direct scram on reactor vessel water level-8 to ensure fuel design limits are not exceeded, the feedwater pump / turbine trip on level 8 is a BWR/6 design feature that j

anticipates a scram and protects the main turbine from damage due to moisture carry-over from the reactor vessel into the main steam lines by j

anticipating a vessel overfill event. The relocation of the level 8 i

feedwater pump / turbine trip function is consistent with the presentation j

in NUREG-1434 since the safety function requirements for a level 8 scram are retained in TS. The requirements associated with these instrumentation functions will be relocated to the UFSAR and will be i

controlled in accordance with 10 CFR 50.59.

i 12.

Existing TS 3/4.3.1 conditions, actions and surveillance requirements for the Flow Biased Simulated Thermal Power - High Scram Function have been relocated to other plant documents. Control of the APRM-FBSTP scram j

function requirements will be relocated to the UFSAR.

The average power range monitor (APRM) flow-biased simulated thermal power (FBSTP) scram is not used or required as a primary scram signal in any Grand Gulf safety analysis. The APRM-FPSTP signal, which is a neutron flux magnitude signal filtered to simulate the thermal flux transfer from fuel to coolant, was chosen for Grand Gulf for first cycle operations, rather than a restrictive critical power ratio (CPR), to protect against events t

with slow thermal or neutron flux increase, for which the thermal flux I

does not significantly lag the neutron flux increase.

The only relevant event for GGNS is loss of feedwater heating (LFWH).

For this event the APRM-FBSTP scram can trip before the neutron flux scram because of its s

-. ~., ~. _,.

. ~.,, -,,

,-,w.

o 48 lower power setpoint. Since cycle 1, however, this scram signal is not i

used by GGNS in this analysis. The initial analysis methodology has been replaced by an approved methodology which examines steady-state conditions before and after a LFWH event. With this analysis there is no i

need for any scram for the event, and thus no need to retain the APRM-FBSTP trip function in the improved TS. The requirements associated with these instrumentation functions will be' relocated to the UFSAR and will j

be controlled in accordance with 10 CFR 50.59.

13. The existing TS 3/4.3.9 conditions, ras, and SRs for the turbine i

overspeed protection system instrumentation have been relocated to other plant documents. The turbine overspeed protection system instrumentation j

is not considered to prevent or mitigate any design basis accident or transient.

i i

Although the design basis accidents and transients include a variety of i

system failures and conditions which might result from turbine missiles i

striking various plant systems and equipment, the system failures and i

plant conditions could be caused by other events as well as turbine failures.

In view of the low likelihood of turbine missiles, this scenario does not constitute a part of the primary success path to prevent or mitigate such design basis accidents and' transients.

Similarly, the turbine overspeed control is not part of an initial 3

condition of a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product i

barrier. The requirements associated with these instrumentation I

functions will be relocated to the UFSAR and will be controlled in accordance with 10 CFR 50.59.

The above relocated requirements relating to installed plant instrumentation are not required to be in the TS under 10 CFR 50.36, and are not required to obviate the possibility of an abnormal situation or event giving rise to an i

immediate threat to the public health and safety.

Further, they do not fall w4hin any of the four criteria set forth in the Commission's Final Policy 2

Statement, discussed in the Introduction above.

In addition, the staff finds that sufficient regulatory controls exist under 10 CFR 50.59. Accordingly, the staff has concluded that these requirements may be relocated from the TS i

to the licensee's TS Bases, or UFSAR, as applicable.

C.

More Restrictive Requirements i

By electing to implement the NVREG-1434 Section 3.3 specifications, the licensee has proposed a number of more restrictive conditions than are allowed by the existing TS. The more significant conditions are the following:

I 1.

Existing TS 3/4.3.2 requires certain inoperable Manual Initiation function channels to be returned to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of their observed inoperability.

Improved TS 3.3.6.1 requires that inoperable Manual Initiation function channel (s) be returned to operable status.within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the affected flow path (s) be isolated. The new requirement for the inoperable manual initiation channel (s)' makes the requirement consistent with the improved TS required actions for all 1

!t i

l 49 other primary containment isolation Manual. Initiation functions. The improved TS changes are acceptable, since isolating the flow path (s) s l

accomplishes the channel (s) safety function.

j 2.

Actions in existing TS LCO 3/4.3.2 do not contain specific requirements i

for when ras or Completion Times are not met; however, the existing TS J

does require entry into LCO 3.0.3.

Existing TS LCO 3.0.3 allows one hour to initiate actions leading to placing the reactor in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> j

and in Mode 4 in the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. -The improved TS Condition K t

specifies guidance if LCO 3.3.6.1 ras or Completion Times are not met.

The improved TS eliminated the one hour allowed to initiate actions l

1eading to the Mode 4 and requires instead that the reactor to be in Mode l

l 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 in the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

i i

3.

Existing TS 3/4.3.2 requirements to take a specified action within one j

hour is changed in the improved TS to " Initiate action to restore channel j

to Operable status.....Immediately." The existing TS requirement would i

appear to provide an hour in which continued operation is acceptable with no action taken, even if the required action could be accomplished sooner. The intent of the required action is more appropriately presented in the improved TS. With the improved TS required action, a significantly more conservative requirement to complete the action as soon as possible is imposed. No longer would the provision to appear to allow continued operation without taking any action for up to one hour.

i i

j This presentation is consistent with the guidance of NUREG-1434.

i 4.

A Note to improved TS 3.3.4.1 RA A.2 precludes the use of the required action to place an inoperable channel in trip if. the instrument channels j

are inoperable due to a trip breaker that will not open.

Placing the channels in the tripped condition will not accomplish the intended j

restoration of the functional capability. The Note is added to prevent this required action from being used in these conditions. Therefore, if l

the instrument channel is not restored in the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> completion time of RA A.I.the LCO Applicability must be exited by reducing power to s 40%

l rated thermal power or removing the associated recirculation pump from

{

service.

j 5.

The Reactor Mode Switch required Modes in improved TS 3.3.2.1 includes Mode 5 applicability with the Reactor mode switch in the shutdown i

position. This was not included in existing TS 3/4.3.6. During this l

condition, the control rod withdrawal blocks are assumed in the safety i

analysis to prevent inadvertent criticality events. Therefore, the functional requirements of the Mode Switch enforced scram must be operable to fulfill the safety analysis assumptions.

6.

Existing TS 3/4.8.4.3 does not include SRs for the verification of the reactor protection system (RPS) under-frequency function time delay setting.

Improved TS SR 3.3.8.2.2 includes the verification of the under-frequency function time delay in the surveillance for Channel 1

Calibration. The under-frequency function time delay is considered in the analysis of the capability of the function to protect the RPS and its inclusion in the SR ensures the proper functioning.

i i

.y

.c, r

y,,

,..,-m---

v.1

d l

a +

a 1

50 7.

Existing TS 3/4.3.2 requires that with less than required minimum operable channels in one trip system for functions other than Main Steam Line (MSL) isolation, the inoperable channel (s) are to be placed in the trip condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Improved TS 3.3.6.1 and 3.3.6.2 have added an additional requirement (RA D.1 for LCO 3.3.6.1 and RA B.1 for LC0 3.3.6.2) when the isolatian capability is not maintained by the automatic functions. The improved TS requires that the isolation capability be restored within one hour. The impact of the improved TS required action is dependent on the instrument logic system. The improved TS is consistent with the existing TS actions for one out of two i

taken twice logic systems. However, for the functions with two out of two taken twice logic, the improved TS is more restrictive in that only one hour is allowed to restore the trip function. The existing TS requirement does not fulfill the intent of restoration of trip systems, with two out two taken twice logic systems and the improved TS Condition B is therefore required.

t 8.

Existing TS 3/4.3.2 requires the Reactor Vessel Water Level - Low, Level 3 trip function for residual heat removal system (RHR) Shutdown Cooling System isolation to be operable during Modes 1 through 5.

Improved TS l

LC0 3.3.6.1 Condition J requires that action be initiated to restore the isolation logic channel to operable status if the RHR Shutdown Cooling l

System is not isolated or if the secondary containment boundary is not established.

9.

The improved TS 3.3.5.1 includes additional low pump discharge flow (bypass) functions of LPCI-A, LPCI-B, LPCS, LPCS-C and HPCS as functions 1.e,1.f, 2.e, 3.f and 3.g that were not included in existing TS 3/4 3.3.

The logic of this instrumentation is important to the proper functioning of the ECCS response to a design basis accident and therefore, the change to improved TS 3.3.5.1 is needed.

10.

Improved TS 3.3.5.1 includes Required Actions F.1 and G.1 that require declaring that the ADS valves inoperable upon the loss of the initiation capability of an initiation function in both trip systems.

Improved TS 3.3.4.1 includes required actions for response to loss of both divisions of the RCIC initiation capability functions. More explicit instruction was included in improved TS 3.3.8.1 for response to loss of the initiation capability of a function for both divisions.

Improved TS Required Action C.1 requires the declaration of inoperability of the associated subsystem within I hour from discovery of the loss of function initiation capability in both divisions. These additions provide clear direction for necessary required actions when the respective conditions apply.

Furthermore, these changes augment the administrative controls program for loss of safety function determination by establishing a practice to enter TS loss of function conditions at the earliest time upon determination of the onset of the condition, 11.

Existing TS 3/4.3.3 does not require the High-Pressure Core Spray (HPCS) injection functions of Drywell Pressure - High and Manual HPCS Initiation to be operable with the reactor vessel water level greater than the Level 8 setpoint coincident with the reactor pressure less than 600 psig. The

I i

(

51 l

i HPCS injection function should be' operable with the reactor vessel pressure less than 600 psig, regardless of the vessel level. -In i

addition, the existing presentation presents the potential for an i

4 unnecessary restriction in restoring normal water level.

If these instruments are inoperable, Specification 3.0.4 will preclude entering the applicability for these instruments, that is, decreasing indicated i

. water level on the wide range instrument to below the Level 8 setpoint while reactor pressure is less than 600 psig. Therefore, this exception to operability is unnecessary, and is therefore deleted.

i 12.

Improved TS 3.3.8.2 includes a condition not-included in the existing.TS l

3/4.8.4.3.

Improved TS Condition D requires all control rods in cells i

l containing fuel assemblies to be inserted, if required actions and associated completion times are not met with one or both inservice' power j

supplies (Condition A) with electric power monitoring assemblies i

inoperable in Mode 4 or 5.

The ' required action for Condition D places the reactor in the least reactive condition and ensures that the safety function of the RPS will not be required.

4 1

13. The improved TS include a new surveillance SR 3.3.4.1.5 to. verify the turbine control valve closure trip and turbine control valve. fast closure i

are not bypassed when the functions are required. Since this a new l

requirement, it is considered a more restrictive change.-

1 l

14.

Improved TS 3.3.2.1 includes new ras for one or more Reactor Mode Switch

- Shutdown Position channels inoperable. With one or more inoperable 1

i channels, the required rod blocks cannot be ensured to be operable. With j

i only one channel, a single failure may prevent the function. Therefore, i

actions are required to fulfill the function of the rod blocks to ensure the required subcritical conditions are maintained.

L 1

15.

Improved TS 3.3.7.1 includes the additional functions low reactor vessel-level and high drywell pressure for actuation of the Control Room Fresh-l Air System.

These functions are currently in the design and assumed to j

initiate the isolation and filtration capabilities of-the system during a-4 design basis accident.

l q

16. An additional PAM function is included in improved TS 3.3.3.1, penetration flow path automatic PCIV position. This function was added i

in accordance wits the NUREG-1434 guideline to include all (1) Type A and j

(2) Category I non-Type A PAM instrumentation.

In addition the number of-i suppression pool water temperature monitoring channels per sector is increased to two, providing TS for single failure monitoring capability.

17.

Existing TS 3/4.3.7.6 channel functional test requirements includes verifying that the SRM count rate is at least 0.7 cps prior to control rod withdrawal. The improved TS SR 3.3.1.2.4 requires the count rate to d

be 3.0 cps, or 0.7 cps with a signal to noise ratio of 2:1. The increased count rate of 3.0 and the signal to noise ratio criteria in the improved TS increases the margin of safety and is a more restrictive 1

requirement.

In addition, the existing TS LC0 3/4 3.7.6 allowance for' no i

minimum count rate during unloading of fuel has been deleted in the f

i i

l*

]

L i

i 52 safety factors.that warrant such an exception.

~

improved TS 3.3.2.1.

Fuel unloading provides no special additional 1

The staff has reviewed the more restrictive requirements and concludes that they result in an enhancement to the improved TS. Therefore, the more restrictive requirements are acceptable.

i D.

Less Restrictive Requirements The licensee in electing to implement the NUREG-1434 Section 3.3 Specifications has adopted a number of less restrictive conditions than are allowed by its existing TS. The more significant conditions are the following:

j 1.

Existing TS 3.3.1 includes Mode 3 and 4 applicability for Reactor Mode

. Switch Shutdown Position and Manual Scram RPS functions.

In Modes 3 and

[

4, all centrol rods are fully inserted and the Reactor Mode Switch Shutdown position control rod withdrawal block (improved TS 3.3.2.1) does not allow any control rod to be withdrawn. Under these conditions, the RPS function is not required to be operable. The actions associated with these functions for Modes 3 and 4 are also removed. Special Operations l

LCO 3.10.3 and LCO 3.10.4 will allow only a single control rod to be j

withdrawn in Mode 3 or 4 by allowing the Reactor Mode Switch to be in the i

i Refuel position. Therefore, the protection provided by the current Mode 3 and 4 RPS operability requirements is adequately provided by the.

l 1

]

improved TS Special Operations LCOs (3.10.3 and 3.10.4).

i 2.

Existing TS 3.3.1 requirements for the removal of RPS shorting links have been removed in the improved TS. The existing TS require that the.

i i

shorting links be removed for the condition of. any control rod withdrawn from a core cell containing one or more fuel assemblies when shutdown margin has not been demonstrated. With the shorting links removed, the i

actuation logic of the APRMs and IRMs are reconfigured to provide a non-coincident scram to prevent inadvertent criticalities.

Shorting link removal in Mode 5 was considered necessary to maintain IRM 3

and APRM operability since the IRMs-and APRMs are not on scale. However, the primary reactivity control functions during refueling are the refueling interlocks and demonstration of shutdown margin. The refueling i

interlocks are required to be operable by LCO 3.9.1 and LCO 3.9.2.

Although the shutdown margin may not have been demonstrated in Mode 5, i

shutdown margin calculations would have been performed and, along with j

procedural compliance for any Core Alterations, would provide assurance i

that adequate shutdown margin is available. Without removing the shorting links, IRM and APRM operability will continue to provide backup 4

with coincident logic' for any significant reactivity excursions. Since 1

the IRM channel high flux scram provides only an uncredited backup in Mode 5, the deletion of the shorting link removal requirement does not j

significantly affect safety.

In addition, the APRM high flux setdown scram will be required to be operable during shutdown margin demonstrations. This TS requirement will i

I

+

y i

j j

i' 53 l

ensure that any time more than one control rod is withdrawn from a fueled cell, neutron flux instrumentation and scram requirements consistent with the requirements for Mode 2 operation are enforced.

l 3.

Improved TS 3.3.1 has modified the existing TS applicability for the j

Reactor Mode Switch Position and Manual Scram functions. The-modification requires only Reactor Mode Switch in the Shutdown Position i

and Manual Scram RPS functions to be operable in Mode 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies. A control rod withdrawn from a core cell containing no fuel assemblies does not affect the reactivity of the core and, therefore, is not required to

)

be operable with the capability to scram.

Provided all rods otherwise remain inserted, the RPS functions serve no purpose and are not. required.

i In this condition, the required shutdown margin (LCO 3.1.1) and the required one-rod-out interlock (LCO 3.9.2) ensure no event requiring RPS will occur. The actions for inoperable equipment in Mode 5 are also i

revised to be consistent with the proposed applicability. Therefore, i

since all control rods are required to be fully inserted during fuel movement (LCO 3.9.3), the proposed applicable conditions cannot be entered while moving fuel. The only possible core alteration is control j

rod withdrawal which is adequately addressed by the proposed action.

i i

4.

Existing TS 3.3.1 includes the Main Steam Line Radiation Monitor (MSLRM)

{

scram function, and existing TS 3.3.2 includes the S LRM isolation j

function. The MSLRM scram function and isolation function have been removed from the improved TS based on the guidelines provided by General-4' Electric NEDO-31400A, " Safety Evaluation For Eliminating the Boiling Water Reactor Main Steam Line Isolation Valve Closure Functint and Scram Function of the Main Steam Line Radiation Monitor." This technical report provided the results of generic evaluations which indicated the j

MSLRMs are unnecessary to ensure compliance with the radiation dose guidelines of 10 CFR Part 100. Additionally, the MSLRM is not credited for a reactor scram initiation for any design basis event. Finally, the.

i reliability assessment of the elimination of the scram function on i

reactivity control failure frequency and core damage frequency indicate a j

net improvement in safety.

i J

With the implementation of the technical report (NED0-31400A) guidelines, j

the operability of trip signals for the mechanical vacuum pump and for cther valves in the design will be controlled outside the TS.

Additionally, the main steam line radiation monitor and off-gas radiation i

monitor alarm setpoints will be standardized at 1.5 times and 2.0 times 1

above the nominal background dose rate to provide an indication of need

]

for a prompt sample of the reactor coolant to determine the possible contamination levels in the coolant and the need for additional I

corrective action. Any significant increases'in the levels of radioactivity in the main steam lines will be expeditiously controlled i

(by procedure) to limit both occupational doses and environmental i

releases.

2 5.

The existing TS 3/4.3.1 completion time for multiple inoperable turbine 4

l control valve or turbine stop valve channels is changed to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from 2 j

l o

i

]

.i 54 L

hours based on operating experience to provide adequate time to reduce l

the power to the levels required in an orderly manner and without challenging the plant systems. Additionally, this time is consistent i

j with the time allowed for similar power reductions'in other proposed j

specifications, e.g. improved TS LCO 3.2.2.

j t

6.

Channel checks in existing'TS 3/4.3.1 for the turbine stop valvesL and for l

4 the turbine stop valves generally consist of only ~a check that the correct indication of valve open or closed position indicator is 4

i provided. There is no actual varying signal to " check" against other j

variables to confirm the channel is indicating properly. Therefore, these'surveillances do not provide the information normally expected from channel checks and their deletion does not significantly impact the reliability of the system.

j j

7.

Existing TS 3/4.3.1 requires-surveillance tests that cannot be performed as required unless the reactor is in the Applicability for the i

1 surveillance. Therefore, a note is added to improved TS SRs 3.3.1.1.2, i

3.3.1.1.10 and 3.3.1.1.12 for high neutron flux detectors to allow entry

]

into the applicable conditions in order to conduct the required tests.

i i

The interval during which the SR must be performed was chosen based on i

i the time for performing the test with a consideration for providing uniform tires to conduct other similar TS required tests.

i 8.

Existing TS 3.3.1 requires operability of the Reactor Vessel Water. Level i

- High, Level 8 function in Mode 1.

The improved TS changed the applicable operational condition to greater than or equal to 25% RTP.

The Level 8 scram is provided to ensure that MCPR is maintained above the i

Safety Limit; however, MCPR is not a concern below 25% RTP due to the-j large inherent margin that ensures the Safety Limit is not exceeded even j'

if a limiting transient occurs, as stated in improved TS Bases for LCO 3.2.2.

Therefore, the LCO Applicability is revised to 25% RTP.

4 9.

Existing TS 3/4.3.1 Action b is separated into two parts which differentiate between inoperable channel conditions that continue to j

provide trip capability and those in which the function loses trip capability. Additional time (from I hour to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) is provided in the i

improved TS Condition A to take corrective action if one or more required channels are inoperable. Additional time (from l' hour to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) is.

j provided in the improved TS Condition B to take corrective action if one or more functions with one or more required channels in both trip systems j

is inoperable, The improved TS restores the RPS to an equivalent l

reliability level as that evaluated in Topical Report NED0-30851-P-A.

In i

i these conditions, a' shutdown or power reduction has been determined not j

warranted. These Required Actions and Completion Time are considered i

i appropriate based on the remainir,g capability to trip, the diversity of I

the sensors available to provide the trip-signals, the low probability of 1

extensive numbers of inoperabilities affecting all diverse functions, and j

the' low probability of an event requiring the initiation of a scram.

i 10.

The existing TS 3.3.2 surveillance test interval (STI) for the SLCS j

initiation logic channel functional test is quarterly on a Staggered Test i

l 1

1

~ ~ -

i.

I 55 Basis. This test has been incorporated into a logic system functional test requirement in the improved TS. The improved TS SR 3.3.6.1.7 STI for the test of the SLCS initiation logic is every 18 months. This function is manually initiated and the testing is very similar to the j

testing of other TS required manual initiation functions which are tested under the existing TS at an 18 month frequency.

The 18 month frequency is based on the need to perform the test under conditions that apply during a plant outage and the potential for an' unplanned trarsient due to l

the loss of the reactor water cleanup system flow when the surveillance is performed with the reactor at power.

In addition, operating experience at other plants that have the same initiation logic system design has shown that these components usually pass the surveillance when performed at the 18 month STI.

i i

11. The existing TS 3.3.2 actions differentiate between channels being

)

inoperable in one or both trip systems. With channels inoperable in both trip systems, existing TS actions do not allow all inoperable channels to-i be placed in the tripped condition even for those cases where this would l

not cause operable systems to isolate. Because of the varied logic in j

isolation actuation systems there is no relatively simple set of actions t

that can be defined to cover all cases of degraded conditions.

Improved TS 3.3.6.1 and 3.3.6.2 have combined the actions for inoperable channels,

)

i independent of whether one or both trip systems are affected. This allows the conservative action of tripping the inoperable channels which

{

is preferable to initiating a shutdown as is currently required'in many 1

cases.

If all channels are not restored or tripped, then the actions of the Table are required as in the current TS.

In addition, a one-hour j

Completion Time is proposed when trip capability is not maintained for restoration or initiation of appropriate action. This time frame is consistent with the one hour provided by LCO 3.0.3 to initiate action for other conditions outside the safety analysis.

i

12. Existing TS Table 3.3.2-1 requires inoperable secondary containment isolation trip function channels to be restored to operable status and l

operation of the standby gas treatment system (SGTS) during core alterations, handling' irradiated fuel in containment, and operations with a potential for draining the reactor vessel. The improved TS replace i

operation of the SGTS with' isolate the affected lines, or declare i

associated valves dampers inoperable and either operate the SGTS or declare the SGTS inoperable. These improved TS options conservatively compensate for the inoperable status of the instrumentation through restoration of the single failure capability or through alternate TS actions for the inoperable equipment or systems. Therefore, the improved TS alternatives to running the SGTS with required function channels

-j inoperable does not impact safety during core alterations, handling irradiated fuel in containment, and operations with a potential for j

draining the reactor vessel.

13.

Existing TS 3.3.3 requires the ADS to be declared inoperable when one or-more actuation channels are inoperable.

Improved TS 3.3.5.1 provides an-

^

option to place the inoperable channels in the trip condition. This j

conservatively compensates for the inoperable status, restores the single i

j j

b

!.o I

56 h

failure capability and provides the required initiation capability of the~

l instrumentation. Therefore, providing this option does not negatively impact safety.

J 14.

Existing TS 3.3.3 requires the instrumentation for HPCS pump suction valve realignment on Reactor Water Storage Tank Level-Low and Suppression i

i Pool Water Level-High to be operable in Modes 4 and 5.

The Modes 4 and 5

?

applicability of the Suppression Pool Water Level-High function was 1

4 deleted in the improved TS.

l t

l The requirements for automatic restoration of the HPCS water source to the suppression pool are dependent on suppression pool availability and the need to require automatic realignment to the pool.

In Modes 4-and 5 l

[

an operable RCIC storage tank can provide an inventory of water to the i

HPCS to adequately minimize the consequences of a vessel drain down event and automatic realignment is unnecessary. Only with insufficient water in the RCIC storage tank is automatic' realignment necessary in Modes 4

}

l and 5.

Therefore, the improved TS Table 3.3.5.1-1 includes in Note (c)

{

an allowance that the requirements of Condition D are not applicable when j

[

the RCIC water tank level. is within limits of SR 3.5.2.2.

j e

i

15. Existing TS 3.3.4.1 LCO requires a reduction to Mode 2 within six hours j

i if remedial actions are not completed for an ATWS Reactor Pump Trip (RPT) i i

function setpoint less conservative than allowed, for operation with less j

i than the minimum required number of channels, or for combinations of one j

or both trip systems inoperable and untripped. The improved TS 3.3.4.2 provides an additional required action allowing the removal of the associated recirculation pump from service. Since this action i

accomplishes the functional purpose of the instrumentation and enables continued operation in a previously approved condition, this change does not have a significant effect un safe operation.

16. An option is provided in improved TS 3.3.4.1 and 3.3.4.2 for one or more inoperable channel (s) to place all inoperable channels in the tripped condition as compared to the existing TS 3/4.3.4 requirements to declare the trip system inoperable. This conservatively compensates for the i

inoperable status, restores the single failure capability and provides j

the required initiation capability of the instrumentation.

In addition, i

the time to meet the action for a single inoperable channel is changed to 14 days from I hour. For this condition the ATWS protection is maintained redundant and not vulnerable to a single failure for either l

pressure or level trip of the recirculation pump, and capable of tripping with the other function but only without additional channel i

j inoperabilities. Therefore, providing this option is commensurate with j

the safety importance of the system and allows for repair of one or more 1

inoperable channels of diverse system sensors configured into redundant trip systems.

i

17. The improved TS 3.3.4.1 required actions for one trip system inoperable are revised to address trip Function capability. This is consistent with i

other specifications which provide appropriate allowed out of service i

times as long as the actuation capability is maintained. Without trip l

i i

l i

4

- -~,,-. -. _

'O Q

l

)

[

57 j

capability prompt remedial measures are required if operation is to centinue. The allowed time for this action is consistent with the existing shutdown requirements established in existing TS LCO 3.0.3 for j

non-compliance with TS.

j

18. The licensee proposed changes to extend the existing TS LCO 3.3.4.1 l

channel functional test monthly surveillance test interval (STI) to once every 92 days, per SR 3.3.4.2.2.

The existing TS STI for the calibration.

f of the trip unit of at least once per 31 days was extended in the l

improved TS to 92 days per SR 3.3.4.2.3.

Additionally, the change to 6 l

hours form 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for performing required channel testing was made.

2 l

Allowed outage times and surveillance test interval TS improvements were accepted by the staff in topical report GENE-770-06-1, February 1991 to provide an acceptable assurance of operability in accordance with a previously conducted reliability analysis.

In a letter of July 29, 1992 I

(GNRO-93/00099), the licensee confirmed the applicability of GENE-770.

1 to the GGNS design with the exception of existing TS 3.3.4.1 functions.

The GENE-770-06-1 changes were approved for GGNS by the staff in l

Amendment Number 105, dated January 6, 1993.

In a letter of October 15, 1993, the licensee certified that the design and reliability assumptions i

for approved changes addressed in topical report GENE-770-06-1 also apply l

to existing TS 3.3.4.1.

19. The existing TS 3/4.3.4.1 required actions for loss of trip capability i

for the EOC RPT instrumentation requires reduction of thermal power to -

i less than 40% of RTP within six hours.

Improved TS 3.3.4.1 provides an 1

optional required action which allows removal of the associated 4

recirculation pump from service in place of a power reduction. Since j

this action accomplishes the functional purpose of reducing the RTP and i,

enables continued operation in a previously approved condition, this change does not have a significant effect on safe operation.

j

20. The existing TS 3.3.6 channel functional test requirements were modified by a Note which allows one hour to conduct the required surveillance-L following entry into the specified conditions. All CFT STI in the i

improved TS are 92 days. This 92 day interval is an increase from the 31 i

day interval in some existing TS LCOs. The 92 day STI is based on a

]

reliability analysis (" Technical Specification Improvement Analysis for BWR Isolation Instrumentation," WEDC-30851-P-A) and setpoint methodology, j

and has been determined sufficient to assure proper operation of the i

required control rod block instrumentation. The Note allows entry into j

the conditions necessary to appropriately perform the surveillance and j

eliminates the need to perform the SR prior to entering the startup mode.

I

21. The proposed change reduces the existing TS 3.3.6 allowable value to turn

]

off the rod pattern controller from greater than equal to 20% reactor thermal power (RTP) to greater than'or equal to 10% RTP in improved TS-l 3.3.2.1.

The RPCS interlocks are initiated below the LPSP and the rod l

withdrawn limiter (RWL) interlocks are initiated above the LPSP. The i

function of the RPCS is to ensure that the peak rod.enthalpy of 280 i

cal /gm will not be exceeded in the event of a control rod drop accident.-

a As shown by the current analysis, this is not a concern at reactor power l

4

l

u-58 greater than 10% power; and, therefore, the RPCS does not place any pattern restrictions on control rod movement above the LPSP. The RPCS is.

demonstrated operable by verifying that the rod pattern controller functions when thermal power is less than the LPSP by selecting and attempting to move an inhibited control rod.

This change is based on a bounding analysis of the control rod drop accident (CRDA) performed by the reload fuel vendor.

The methodology employed in performing the bounding CRDA analysis supporting this change and the analyses performed to support reload cycles 2 through 7 is defined in XN-NF-80-19(P)(A),. " Exxon Nuclear Methodology for Boiling l

Water Reactors - Neutronic Methods for Design and Analysis", Volume 1 and l

Supplement 3 Exxon Nuclear Company Inc., Richland, WA, March 1983. This document is approved for use in establishing Grand Gulf cycle specific operating limits as described in current GGNS Technical Specification 2

6.9.1.11.

The acceptance that the CRDA was self limiting above 10% RPT l

was provided in Amendment 73.

22. Existing TS SR 3/4.1.4.1 actions require immediate return of control rod (s) to the position prior to withdrawal when the main turbine bypass valves are not fully closed and thermal power is greater than the low 1

power setpoint of the rod pattern control system. The improved TS changed this requirement to immediately suspend control rod withdrawal.

It is not always possible to return the control rods to the position

)

prior to the control rod withdrawal if it is not known when the bypass valves were opened. Additionally, reinserting the control rods may not be necessary for continued safe operation. Therefore, the required 4

action is limited to suspend control rod withdrawal which meets the LC0 requirement for an inoperable rod pattern control system. In addition, the existing TS requirement for a second licensed operator, to verify control rod withdrawal is prevented, has been deleted. This is adequately addressed by LC0 3.3.2.1 RA's.

23. The existing TS 3.3.7.1 Control Room Ventilation Radiation Monitor 4

applicability is reduced to Modes 1, 2, and 3, and during operations with a potential for draining the reactor vessel in the improved TS 3.3.7.1.

This change limits the applicability of the requirements for the system to during those operations which have potential to create a need for the system to operate. The omitted conditions are not considered initiators for events which require the system and therefore, the change does not impact safety.

24. Existing TS 3.3.7.1 allows one hour to initiate and maintain nperation of at least one control room emergency filtration system in the isolation i

mode of operation when both of the required monitors in a trip system are inoperable. The improved TS extends the time to two hours (one hour to declare the systems inoperable and another hour to put one in operation).

This time provides the operator with time for restoration or to place the subsystem in operation, without posing a significant change in the operation of the plant.

4 i

l.

b i

59

25. The existing TS 3.3.7.4 Allowed Outage Time (A0T) for inoperable remote shutdown system instrumentation and controls is extended to 30 days. The system is not required to respond to any design basis accident evaluated l-in the safety analysis, but is provided to comply with GDC-19. The i

Specification is retained as improved TS 3.3.3.2 only as a significant contributor to risk reduction, and extending the A0T does not have a i

significant impact on that contribution.

1 1

26.

Existing TS 3.3.7.4 required a Channel Check on the required remote j

shutdown system instrumentation.

Improved TS 3.3.3.2 requires the Channel check to be performed only when the channel is normally i

energized. Operating history has shown that the equipment passes the i

calibration surveillance conducted on an ~18 month frequency. Therefore, the Channel Check provides limited additional assurance of the j

capabilities of the system. Additionally, performance of the Channel Check currently requires energizing a normally dormant portion of the i

- system specifically to perform this surveillance. Therefore, the Channel l

Check is proposed to only be required when the system is normally.

3 energized.

l

27. The existing TS 3.3.7.5 requirements for post accident monitoring instrumentation in Mode 3 were deleted in improved TS 3.3.3.1.

Accident monitoring is provided to assist in the diagnosis and preplanned actions i

required to mitigate design basis accidents which are assumed to occur in i

Modes 1 and 2.

The probability of an event in Mode 3 that would require j

accident monitoring instrumentation is sufficiently low that the accident i

monitors are not required in these Modes.

28. The existing TS 3.3.7.5 A0T for one channel of inoperable accident monitoring instrumentation is extended from 7 days to 30 days and the A0T for two channels of inoperable accident monitoring instrumentation is extended from 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to 7 days.

Existing TS required actions for one 1

accident monitoring channel inoperable in one or more functions for more l-than 30 days is revised from requiring a' shutdown to requiring a special report in accordance with the Bases of improved TS 3.3.3.1.

3 Due to the passive function of accident monitoring instrumentation and l

the operator's ability to respond to an accident utilizing alternate i

instruments and methods for monitoring, it is not appropriate to impose stringent out of service times.

In some instances the existing A0T for l

these monitoring instruments is shorter than the A0T for the system which is needed to maintain the monitored parameter within limits.

Further, 1

because of the likelihood that the instrumentation can be repaired in the allowed time, indefinite operation with inoperable functions can be supported since the specified TS report will ensure alternate monitoring i

methods are available while the accident monitors are inoperable.

j'

29. The existing TS 3.3.7.5 channel calibration surveillance interval on the j

hydrogen analyzers and monitors is changed from monthly to once per 92 i

days. The only time dependent component of the current setpoint uncertainty calculation (recorder drift) is conservative for a 92 day j

calibration interval, and increasing the calibration interval to 92 days i

1 l

j'

[

O' I

1 60 i

has no' effect on the calculated total loop uncertainty'for the hydrogen -

J analyzer and monitor.

In addition, the. channel check continues to be i

required once per 31 days. Therefore, given the redundant l

instrumentation and the existing channel check the proposed surveillance

)

interval increase is acceptable.

j

30. A Note was added in improved TS 3.3.1.2 for the source _ range monitor i

surveillance requirements. The Note deletes the count rate verification

)

surveillance requirements when less than or equal to four fuel assemblies i

j are adjacent to the SRM and no other fuel. assemblies in the core quadrant are present. The SRMs are provided to monitor changes in reactivity that.

may lead to unintentional criticality. With four or-less fuel-assemblies-l l

loaded around each SRM and no other fuel assemblies. in the associated i

quadrant, the configuration will not be critical even with a control rod l

l withdrawn. Therefore, the count rate requirement _ for the monitor is less i

stringent.

31. The existing TS 3.3.7.6 and existing TS 3.9.2 shorting link removal l

requirements have been removed from the source range monitor l

l requirements improved TS 3.3.1.2.

In the existing TS, removal of the shorting links was required when any control rod was withdrawn-from a core cell containing one or more fuel assemblies when shutdown margin had l

not been demonstrated.

Shortin link removal was considered necessary in Mode 5 to maintain IRM

)

l operabi ity since the IRMs are not on scale and only the SRMs (with the shorting links removed) provided the necessary input for a high flux

~

scram through the IRM channels. However, the primary reactivity control functions during refueling are the refueling interlocks and the shutdown margin. The refueling interlocks are required operable'by improved TS LCO 3.9.1 and LCO 3.9.2.

Although shutdown margin may not yet have been demonstrated in Mode 5, shutdown margin calculations would have been performed and, along with procedural compliance for any core alterations, would indicate that adequate shutdown margin is available. Without removing the shorting links, IRM operability will continue to provide backup for the credited functions for any significant reactivity i

excursions. Since the IRM channel high flux scram provides only an uncredited backup in Mode 5, the deletion of the shorting link removal requirement does not significantly affect safety.

l 32.

In improved TS 3.3.1.2, the SRM operability requirements in Mode 5 have been modified.

In Mode 5, during a spiral offload or reload, a SRM

}

outside the fueled region will no longer be required to be operable, i

since it is not normally capable of providing adequate monitoring of neutron flux in the fueled region of the core. However, the SRM detector in the fueled region must be operable, and this single detector is sufficient to effectively monitor core reactivity changes with on-scale monitoring instruments.

I 33.

Existing TS 3.3.7.6 and TS 3.9.2 SRs for SRMs included surveillance i

performance prior to core alterations, prior to control rod withdrawal, and prior to moving the reactor mode switch. -Improved TS 3.3.1.2 deletes i

)

i

-r<r-

+-

n.

e s

n---we-m-e, v.- m s

l.-

~

i 1

i 61 i

these requirements. This deletion removes an unnecessary additional

{

. performance of a surveillance because moving the reactor mode switch, withdrawing control rods, and performing core alterations do not impact the ability of the SRMs to monitor core reactivity changes. The SRM surveillance tests are now required to be performed periodically while in i

Modes 2, 3, and 4.

The required testing ensures core reactivity changes are effectively monitored with'on scale monitoring instruments.

Therefore, requiring an additional surveillance prior to core alterations, control rod withdrawal, and moving the reactor mode switch l

1s unnecessary.

34.

In existing TS 3.3.7.6 while in Mode 2 with more than one inoperable SRM channel, the TS require a shutdown, per LCO 3.0.3.

The improved TS have i

added Condition B which requires immediate suspension of control rod withdrawal with all required SRM channels inoperable. This change will

)

allow continued operation with no operable channels if all positive l

reactivity changes due to control rod withdrawal are suspended. The existing TS requirement to shutdown per LCO 3.0.3-is unnecessarily restrictive and does not allow concentration of efforts to repair inoperable instrument channels.

Improved TS Action A.1 still applies and allows four hours to restore monitoring capability prior to requiring control rod insertion. Since the prevention and mitigation of prompt-reactivity excursions during refueling and low power operation are provided by improved TS 3.9.1,'3.3.1.1 and 3.3.2.1,~the action to suspend control rod withdrawal as compensation to loss of monitoring capability is sufficient to avert an immediate threat to public health and safety.

35. Existing TS 3.3.7.6 requirements to perform a channel functional test of the SRMs in Mode 5 are deleted since in the improved TS the SRMs only provide the operators with information on neutron flux levels at low power levels and during shutdown conditions, including core alterations and do not provide any trip functions. Other related functions, such as the control rod block function of existing TS 3.3.6 requirements for the SRMs are relocated from the TS to a licensee controlled document based on an evaluation of the final policy statement for TS. The functional capability of the SRM indicators are adequately tested during Mode 5 l

operations by improved TS SR 3.3.1.2.1 which provides for a channel check at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with adequate count rate (per SR 3.3.1.2.4) and by SR 3.3.1.2.7 which provides for a channel calibration at least i

once per 18 months. Furthermore, there are not additional functional i

requirements that need to be met for the SRM channels while in MODE 5.

In Modes 2, 3, or 4, performance of a channel functional test at least once per 31 days provides sufficient assurance of the functional capability of the required SRM channels'in Mode 5.

36. Existing TS 3.3.8 Actions do not provide an option to place all inoperable channels in the tripped condition when the number of operable channels is less than required by the minimum operable channels requirement.

If more than one channel is inoperable, the existing-TS requires declaring the associated subsystem inoperable.

Improved TS 3.3.6.3 and 3.3.6.4 provide the option to place. inoperable channels in trip. This conservatively compensates for the inoperable status,

.~

e l

i 62 restores the single failure capability and provides the required initiation capability of the instrumentation.

37. The existing TS require LCO 3.8.4.3, Reactor Protection System Electric Power Monitoring, to be applicable during Modes 4 and 5.

Improved TS 3.3.8.2 requires RPS electric power monitoring instrumentation to be applicable in Modes 4 and 5 with any control rod withdrawn from core cells containing fuel assemblies. With no control rods withdrawn from core cells containing fuel assemblies, there is no need to require the RPS scram function to be operable and, therefore, no need for the LCO requirement.

38. The existing TS require the restoration of the RPS electric power monitoring channel to operable status within 30 minutes. The improved TS 1

increases the restoration time to one hour. The improved TS 3.3.8.2 thirty minute time extension is minimal considering the likelihood of a condition of degraded grid voltage concurrent with a design basis accident requiring actuation of the RPS. The one hour time is in agreement with the time allowed for all other required instrument functions under similar conditions and does not present a significant change in the operation of the plant.

39. Existing TS 3.3.3 requires the ADS function to be operable at reactor pressures greater than or equal to 135 psig. The improved TS increase this requirement to greater than 150 psig to maintain consistency of the operability requirements for all ECCS and RCIC equipment. Small break Loss of Coolant Accidents (LOCAs) are not assumed or analyzed to occur at low pressures (i.e., between 135 psig and 150 psig). The ADS lowers reactor pressure sufficiently during a 'small break LOCA so that the low.

pressure core injection and core spray systems can provide makeup to mitigate high pressure accidents. Since these low pressure systems begin to inject water into the reactor pressure vessel at pressures well above 150 psig there is no safety significance in ADS not being operable between 135 and 150 psig.

40.

Improved TS SR 3.3.2.1.9 allows control rods to be bypassed and ~

repositioned under the direction of a second licensed operator or other qualified member of the technical staff. This allowance is currently only applicable (per existing TS 3.1.4.2 Action b) for insertion of the bypassed control rod.

Existing special test exception TS 3.10.2 allows a bypassed control rod to be withdrawn to its prior position to support special tests, including scram time tests.

The improved TS expand this flexibility to allow the control rod to be bypassed with the positioning of the control rods controlled by other specifications as discussed below.

LCO 3.1.4 does not allow slow or stuck control rods to occupy adjacent locations. LC0 3.1.6 requires operable control rods to be in compliance i

with the Banked Position Withdrawal Sequence (BPWS) analysis while LCO 3.1.3 Condition D requires inoperable control rods to either comply with the BPWS analysis or to be separated by at least two core cells to ensure that the control rods remain within the patterns assumed for the CRDA

j e

63 analysis. This ensures that the control rods remain within the patterns assumed for the CRDA analysis, which is only of concern below the Low Power Setpoint (LPSP) of the RPC system. With respect to the Rod i

Withdrawal Error (RWE) accident, which is only applicable above the LPSP of the RPC system, adequate separation of control rods is assured by the -

4 j

thermal operating limits.

1 As stated above, the improved TS expand the flexibility for bypassing and I

moving bypassed control rods.

Implicit in the requirement that movement-i of the bypassed control rods be performed under the direction of a second j

licensed operator or other qualified member of the staff is that-the positioning be in conformance with applicable safety analyses. When operating below the LPSP of the RPC system, the applicable analysis is -

)

i the CRDA. Compliance with this analysis is ensured by conformance with i

i the generic BPWS analysis or a specific BPWS analysis for the evolution.

1 Similarly, when operating above the LPSP of the RPC system, the 1

applicable analysis is the RWE analysis. Movement of bypassed control l

rods must be in conformance with the generic RWE analysis or with a special analysis to ensure that the conclusions of the RWE analysis i

remains supported. These controls ensure that positioning and movement of bypassed control rods remain within the bounds of previous analyses.

}

41.

Existing TS 3.3.1, includes an exception to SR 4.0.4 for IRM Channel i

functional testing.

Improved TS 3.3.1.1 provides an additional exception i

i to SR 4.0.4 for the APRMs. This exception was added since the APRMs are required in Mode 2, although the required surveillance cannot be j.

performed in Mode 1 (prior to entry in the applicable Mode 2 from Mode 1) without utilizing jumpers or lifted leads. Use of these devices is not recommended since errors in their use may significantly increase the r

i probability of a reactor transient or event which is a precursor to a i

previously analyzed accident. Therefore, this exception allows 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> j

to conduct SR 3.3.1.1.3 after entering Mode 2.from Mode 1.

This change j

is consistent with NUREG-1434.

j 42.

Existing TS 3.3.1 IRM Neutron Flux-High and APRM Neutron Flux-High j

surveillances prior to startup channel functional test requirements have been deleted in improved TS 3.3.1.1.

The improved TS requires these a

surveillances to be performed periodically while in Modes 2, 3, and 4.

This frequency has been determined to be sufficient verification that the i

APRM logic is properly functioning. Moving the reactor mode switch, i

withdrawing control rods, and performing core alterations do not impact i

the ability of the monitors to perform their required function.

Therefore, requiring performance of an additional surveillance prior to one of these events is unnecessary.

43.

Existing TS 3.3.2 requires that, with less than the minimum required i

manual initiation function channels operable for secondary containment L

isolation, the inoperable channel (s) must be restored to operable status within eight hours or the system isolation valves must be closed within one hour.

Improved TS LCO 3.3.6.2 allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to place the

~

inoperable manual initiation function channel (s) in the trip position.

4 If the manual initiation function channel (s) are not restored to operable l

i

.j 4

l 1

64 status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the improvod TS requires the affected flow paths-l to be isolated within one hour or other alternate actions taken.

The time allowed by the improved TS to restore the. manual initiation function channels to operable status is consistent with the required actions for other manual initiation functions for system isolation and provides time for adequate preparation to accommodate the effects of isolating the penetrations.

In addition, the improved TS changes are acceptable since the manual initiation function is not assumed in any accident or transient analysis lin the UFSAR and alternate conservative i

compensatory actions (LCO 3.3.6.2) are provided if the penetrations i

cannot be isolated by the manual initiation function.

44. Existing TS Table 3.3.2-1 requires that, with less than the required minimus MSL isolation function channel (s) operable (except for Reactor Vessel Water Level - Level 1), the reactor is to be in the startup mode and the associated isolation valves closed within six hours.. With.less-than required minimum Reactor Vessel Water Level - Level 1 function channels operable, the existing TS requires the reactor to be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and cold shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The improved TS removes the requirement to place the reactor in the startup Mode and increases the time from 6 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for isolating the associated MSL for all functions except' Main Steam Line Pressure -Low.

The improved TS also makes the required action for the Reactor Vessel Water Level - Level 1 function the same as the other MSL-isolation functions.

In addition, the improved TS requirement for less than required minimum Main Steam Line Pressure - Low function channels operable, removes the requirement to close the associated isolation valves and retains only the existing TS action to be in the startup Mode within six hours.

The required action to isolate all associated MSLs with the functions-inoperable necessitates being in Mode 2 (startup) to avoid a scram. The requirement to be in Mode 2 is therefore implicit and is deleted. The additional six hours to isolate the associated MSL allows for more orderly power reduction. Since 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> are otherwise provided to exit the applicable modes per LCO 3.0.3, this short extension is still well within reliability or availability assumptions for redundant TS functions to provide sufficient protection from design basis accidents. With less than the required Reactor Vessel Water Level -Level I channels operable, it is not necessary to require a shutdown of the unit if the conditions only affect the isolation logic for one MSL.

In these cases, the improved TS allows isolation of the affected line.

This returns the system to a status where it can perform the remainder of its isolation function, therefore permitting continued operation, although possibly at a reduced power level.

The existing TS requirement to close the associated isolation valves within six hours with less than required minimum MSL Pressure-Low function channels operable is unnecessary since the same action in the improved TS requires the reactor to be in startup within six hours. The 4

~_ _

i a

x 65 L

MSL Pressure-Low isolation function is not required in the startup Mode i

by either the existing TS or improved TS.

i j

45. Existing TS Table 3.3.2-1 includes the requirement that, with less than j

the required minimum manual initiation function channels for isolation 1

operable, the inoperable channel (s) must be restored to operable status i

within eight hours or the plant must be shutdown.

Improved TS Table -

l 3.3.6.1-1 allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to isolate the affected flow paths or to shut l

down the plant.

The time allowed by the improved TS to isolate the affected flow paths, and restore the inoperable Manual Initiation function channel (s) to operable status, is consistent with the required actions for other manual I

initiation functions for isolation and provides time for adequate preparation to accommodate the effects of isolating the penetrations.

In i

addition, the improved TS changes are acceptable, since the manual 1

initiation function is not assumed in any accident or transient analysis j

in the UFSAR.

i l

46. Existing TS 3.3.2 includes requirements for the SLCS Initiation function I

for Reactor Water Cleanup System Isolation in Mode 5 with any control rod i

l withdrawn.

Improved TS 3.3.6.1 remove the requirement for the SLCS Initiation function during Mode 5 with any control rod withdrawn.- The j

SLCS Initiation trip function is not required in Mode 5, since only a l

single control rod can be withdrawn and adequate shutdown margin prevents j

criticality under these conditions.

i j

47. Existing TS 3.3.2 allows one of two reactor water level 3 trip systems i

for the isolation of the RHR system to be inoperable for up to 14 days during Modes 4 and 5.

Improved TS 3.3.6.1 requires only one trip system to be operable in Modes 4 and 5, if the integr ty of the RHR shutdown i

cooling system is maintained. Since system i slation on low water level i

is provided in Modes 4 and 5 to mitigate a vessel draining event, an intact RHR system fulfills the function of one trip system. Therefore, l

only one operable trip system on Level 3 is required to be available to j

mitigate a vessel draindown event, if the RHR system is intact.

i 1

48. The existing TS 3.3.2 isolation actuation instrumentation requirements are based on the requirements for the associated valves which receive 1

isolation signals from the instrumentation. As a result, low reactor water level; ventilation exhaust high radiation; and high containment

{

pressure signals for primary containment and secondary containment isolation are required to be cperable during Core Alterations, while j

handling irradiated fuel in the primary or secondary containment, and j

during operations with a Potential for Draining the Reactor Vessel (OPDRVs). However, these requirements do not take into consideration the 4

ability of the specific event sequences during these operating conditions i

to generate an isolation signal. The improved TS 3.3.6.1 action requirement changes are discussed below for each of the isolation 3

signals.

I i

l:

E i

.~.

l.

i 66 i

i An event involving Core Alterations or fuel handling cannot create a low j

reactor water level condition and, as discussed in the GGNS UFSAR,' GGNS does not have secondary containment bypass leakage-paths. Thus, low.

1 i

reactor water level instrumentation is not assumed 'to mitigate events postulated to occur in any of these plant operating conditions.

Because an event involving Core Alterations, fuel handling or OPDRVs can i

create an increase in ventilation exhaust ' radiation, these instruments I

will continue to be required operable in these plant conditions, consistent with the existing TS requirements.

Manual isolation signals for primary containment and secondary I

j containment ventilation exhaust high radiation, have requirements to be 1

operable during Core Alterations, while handling irradiated fuel in the j

primary or secondary containment, and during OPDRVs. Therefore, the j

improved TS applicability requirements are changed to establish the j

manual function capability during the conditions for which the manual 1

function supports the automatic isolation signals.

1 i

The proposed changes are acceptable on the basis that the valves which 1-isolate secondary containment and those primary containment isolation 1

valves which support the secondary containment. isolation function will J

continue to be required to be operable by LCOs in Section 3.6, improved j

TS LCO 3.6.4.2 and LCO 3.6.1.3, respectively.

In addition, the associated isolation instrumentation is required to be-operable during those plant conditions in which the instrumentation is assumed to j

9enerate an isolation signal in response to the postulated event.

49. The existing TS 3/4.3.2 for primary containment and secondary containment isolation instrumentation response time testing corresponding to the j

required diesel start time is deleted. These changes are based on 9uidance provided in Generic Letter.93-05. Line-Item. Technical l

Specifications Improvements to Reduce Surveillance Requirements for Testing During Power Operation, September 1993.

i As described in the current Technical Specification bases, with the exception of Main Steam Isolation Valves (MSIVs), individual sensor j

)

response timer te the response times of the logic systems to which the sensors are connected are not addressed in the safety analysis.. For D.C.

operated valves, a 3 second delay is assumed before the valve starts to move. For A.C. operated valves, it is assumed that the A.C. power supply is lost and is restored by startup of the emergency diesel generators.

l In this event, a time of 10 seconds is assumed before the valve starts to i

}

move.

In addition to the pipe break, the failure of the D.C. operated i

valve is assumed; thus the signal delay (sensor response) is concurrent T

with the 10 second diesel startup. The safety analysis considers an 1-allowable inventory loss in each case which in turn determines the valve speed in conjunction with the 10 second delay.

It follows that checking i

the valve speeds and the 10 second time for emergency power establishment will establish the response time for the isolation functions. Thus, the signal delay (sensor response) is concurrent with the 10 second diesel startup.

Since typical response times are measured in fractions of a 7

l

.i

~

67 i

second, the chance is remote that a channel's response would degrade to j.

the point where it exceeds the 10-second diesel start time without a noticeable failure.

}

50. The existing TS 3/4.3.7.1 requirements to restore the inoperable channel I

to operable status within seven days or within the next six hours, and initiate and maintain operation of at least one control room emergency i

filtration system in the high radiation mode for one required monitor in i

a trip system inoperable, are deleted from improved TS 3.3.7.1.

This change omits the requirement to restore the inoperable channel to e

1 operable status within an identified time frame. The channel must be j

placed in the tripped condition within a few hours, and is therefore, j

fulfilling the required safety function. There is also no need to place i

the system in service since a valid signal will continue to automatically j

actuate the system. Therefore, the safety function is maintained.

t c

{

51.

Existing TS LCO 3.3.2 Action b.1 requires, that if placing inoperable i

isolation actuation channels in the tripped condition would cause an isolation, the inoperable channel (s) shall be restored to operable status I

within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, or the required actions for the affected trip function' i

shall be taken.

If placing the inoperable channel (s) in the trip condition would not cause isolation, the existing TS requires that t

l inoperable channel (s) be placed in the trip position within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for i

trip functions common to the RPS instrumentation and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for 1

trip channels not common to the RPS instrumentation.

Improved TS 3.3.6.1 l

1 eliminates the difference in Completion Time for the restoration of l

inoperable trip channels between the channels that placing in the trip position would result in a trip and those that would not.

Improved TS ras A.1 and A.2 require placing the inoperable channels in the trip

{-

position within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for channels common to the RPS and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for channels not common to the RPS.

If the inoperable trip channels are not 1

(

placed in the trip position (or restored to operable status) within the required Completion Time, the improved TS requires entry into Table l

3.3.6.1 referenced conditions. The improved TS Completion Time to i

restore inoperable trip channels to operable status is consistent with i

NEDC-31677-P-A, Technical Specification Improvement Analysis for BWR Isolation Action Instrumentation, June 1989 and NEDC-308512, Supplement 2, Technical specification Improvement Analysis for BWR Isolation i

Instrumentation Common to RPS and ECCS Instrumentation, March 1989. The i

12 and 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> out of service times are acceptable the based on the

]

diversity of sensors and the redundancy of isolation design.

1 52.

Existing TS 3.3.5 pressure at which the RCIC actuation instrumentation is l

required operable is increased to 150 psig to provide consistency of the I

operability requirements with the ECCS. Small break loss of coolant accidents are assumed, or analyzed to occur at low pressures (i.e.,

]

between 135 and 150 psig). The RCIC is designed to operate to maintain the level at high pressure. The low pressure core injection and core i

spray systems can begin to inject water into the reactor pressure vessel i

at pressures well above 150 psig (e.g., 225 and 289 psid). Therefore, there is.no safety significance in the RCIC not being operable between 4-135 and 150 psig.

i 1

- - +

e w-

,,y r

m ----.

ce..

-,v,,,-r w-,-w%w.-

w w

2 o.

i 4

68 j

53.. Existing TS 3.3.5 Action 50 does not provide an option to' place all inoperable channels in the tripped condition when the number of operable channels is less than required by the Minimum operable Channels requirement. This option is provided by the improved TS, Condition B.2, i

to place inoperable channels in the tripped condition. This i

conservatively compensates for.the inoperable status:and restores the single failure capability and provides the required initiation capability of the instrumentation.

54. Existing TS 3.3.6 Channel Calibration STI for the rod block functions of i

low power setpoint and high power setpoint were changed from every 92 j

days (Q) to every 184 days based on plant specific setpoint methodology.

i This change of frequency does not include the trip unit calibration which -

]

will continue to be performed on a 92 day frequency.

Tb1 above less restrictive requirements have been reviewed by the staff and i

i hue been found to be acceptable, because they do not present a significant j

safety question in the operation of the plant. The TS requirements that i

j remain are consistent with current licensing practices,' operating' experience i

j and plant accident and transient. analyses, and provide reasonable assurance i

that the public health and safety will be protected.

I i

l 3.4 Reactor Coolant System i

A.

Significant Administrative Changes In accordance with the guidance in the Final Policy Statement, the licensee i

has proposed administrative ~ changes to the existing technical specification l

{

(TS) to bring them into conformance with the improved TS. These changes are i

as follows-i 1.

Existing TS 2.2.1, " Limiting Safety System Settings," existing TS 3.2.1, j

" Average Planar Linear Heat Generation Rate," existing TS 3/4.4.1,

" Recirculation Loops," and 3/4.4.1.3, " Recirculation Loop Flow,"

l applicability requirements provide notations to the operator that Special

~

Test Exception 3.10.4 has parallel applicability requirements. The format of the improved TS eliminates the LCO " cross references."

i Improved TS LCO 3.0.7 adequately prescribes that the use of the Special i

Operations LCOs is optional without the need for such references to establish appropriate requirements. Therefore, existing LCO cross l

references are retained in the improved TS as LCO 3.0.7 and the removal of the LC0 notation is acceptable.

i 2.

The operation of the Safety / Relief Valves (S/RVs) is proposed to be included in the system functional tests of improved TS SR 3.4.4.2 and SR 3.4.4.3.

The existing TS definition of Logic System Functional Test

]

(LSFT) includes provisions to test the actuated device, but is been revised in the improved TS to exclude actual operation of the equipment.

The actuated device is to be tested as part of the system functional test. Deleting the actuated device from the definition of LSFT eliminates the confusion as to whether a previously performed LSFT is l

2 i

- ~

- - - - ~ - -.. - _ - - _

s v

l 69 rendered invalid if the final actuated device is discovered to be i

inoperable as a consequence of another surveillance (e.g., valve.

cycling). Therefore, including the SR/V as part of the above identified l

improved TS surveillance requirements effectively retains required operation of the equipment within the improved TS.

i 3.

Performance requirements of the Primary Coolant Specific Activity Sample j

and Analysis Program in existing TS Table 4.4.5-1 are reformatted within l

Required Actions A.1 and 8.1 of improved TS 3.4.8,- reactor coolant system (RCS) Specific Activity. The performance requirements are effectively i

j retained within the improved TS and are therefore acceptable.

J l

l-4.

The existing TS 3/4.4.6.1 Action to restore the temperature and/or i

2 pressure to "within 30 minutes" is proposed to be revised to " initiate

-l action to restore the LCO limit immediately" for Modes 4 and 5.

The intent is to return the reactor coolant pressure boundary to a condition l

that is verified to be acceptable by stress analysis. The existing t

requirement can provide a half hour in which pressure and temperature j

i requirements could exceed the limits, even if capable of.being returned j

to within limits more promptly. The accepted action is more appropriately presented in improved TS 3.4.11 Required Action C.I.

i i

5.

Existing TS SR 4.4.6.1.3.b requires verification of the vessel flange and head flange temperature 30 minutes prior to tensioning of the head bolting studs and every 30 minutes thereafter.

Improved TS SR 3.4.11.5 j

deletes the requirement to verify temperature 30 minutes. prior to i

i tensioning the bolting studs. Since this SR must be current prior to tensioning the Reactor Pressure Vessel (RPV) head bolts in accordance with SR 3.0.4, the requirements are essentially unchanged. 'Therefore,

.1 the improved TS is acceptable.

4 6.

Existing TS SR 4.4.6.1.4 specifies removal and examination of reactor vessel material specimens to determine changes in the RPV material properties as a function of time and thermal power.

Existing TS-j 4.4.6.1.5 specifies that PT limit curves are valid for' up to 10 effective g

full power years and that a scheduled updating of the curves is required to be conducted before the limitation is exceeded. The results of the examination are used to adjust the curves of TS figure 3.4.6.1-1 as j

required by 10 CFR Part 50 Appendix H.

Since the regulations require compliance and the withdrawal schedule cannot be changed without prior approval, as necessary, pursuant to 10 CFR 50.59 the recitation of this j

regulatory requirement in the TS is unnecessary.

7.

Improved TS 3.4.11 Condition A. is modified by a Note that clarifies to the operator that once the Pressure-Temperature (PT) limits are exceeded i

in Modes 1, 2, or 3, the required engineering evaluation is to continue to its completion regardless of when the required action is exited.

Thus, the existing TS requirement is maintained within the improved TS and the proposed change is acceptable.

a.

i 8.

A Note to improved TS SRs 3.4.11.6 and 3.4.11.7 is provided to clarify l

the current intent of allowing entry into the applicable Modes without 4

l

~

i*

70 i

having performed surveillance requirements to verify reactor vessel flange and head temperatures. This requirement is only performed during-

'the specified conditions, and the clarification is consistent with

. current TS requirements. Therefore,-the change is acceptable.

i 9.

Existing TS 3.4.9.1 Action a. requires a periodic surveillance to verify i

once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that an alternate decay heat removal method is available. This requirement.is moot since the.same action requires the i

reactor to be in Mode 4 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Once in Mode 4, existing TS.

3.4.9.2, Residual. Heat Removal (RHR) Cold Shutdown -(improved-TS 3.4~ 10,

" Residual Heat Removal (RHR) Shutdown Cooling System - Cold. Shutdown")

requires periodic verification of the availability of an alternate decay i

heat removal method. Since the surveillance frequency in existing TS j

.3.4.9.1, Action a. is essentially. retained in the improved TS, the change is acceptable, i

10. A Note, " Separate condition entry is allowed for each RHR Shutdown-l Cooling Subsystem," is proposed to be'added to e'xisting TS 3.4.9.1 and j

3.4.9.2.

This' Note provides explicit instructions for proper application.

j of the actions for improved TS compliance.

In conjunction with the j

improved Specification 1.3

" Completion Times," this Note provides-3 direction consistent with the existing TS, and the change is therefore I

j acceptable.

11.

In a letter of September 23, 1993 (GNRO-93/00118), the licensee proposed j

eliminating SR 4.4.1.2.1, which takes exception to the provisions of Specification 4.0.4 if the diffuser to lower plenum differential pressure j

meets established limits. The staff approved this change in Amendment l

110, on January 4, 1994.

12. Existing TS 3/4.4.1.1 does not require action if no recirculation loop is i

operating; however, existing TS LCO 3.0.3 directs the plant to be placed

}

in a mode for which the LCO no longer applies; Improved'TS 3.4.1, Condition F, is added to clarify the required operator action if no a

l recirculation loop pumps is operating while the plant is in Mode 2.

This change is acceptable because the existing -TS acticn requirements are the 1

same as those in the improved TS Condition F.

i 13.

In a letter of August II, 1993 (GNRO-93/00097), the licensee proposed TS l

changes to the guidance for determining whether the tail pipe pressure indication system for the relief valve is operable. This change is t

consistent with the guidance of NUREG-1434 on the loss of trip function l

redundancy. The staff accepted the proposed changes in License Amendment j.

118, dated February 16, 1995.

14. Thermal stresses on vessel components during recirculation loop startups are dependent on the temperature difference between the idle loop coolant j

i and the RPV coolant.

Improved TS SRs 3.4 11.4 and 3.4.11.9 ensure the.

temperature difference between any loop to be started and the RPV coolant is acceptable. A separate LCO requirement to monitor the temperature difference between an idle loop and an operating loop in existing TS 3.4.1.4 and 3.4.1.1 are redundant to the loop-to-coolant requirements of

[

l.

I i

i i.

71 SRs in LCO 3.4.11.

However, the loop-to-coolant temperature check may j

use the operating loop temperature as representative of " coolant j

temperature."

The above changes result in the same limits as the current requirements, or they represent an enhanced presentation of the existing TS intent.

Accordingly, the improved TS changes are purely administrative and are acceptable.

B.

Relocated Requirements l

In accordance with the guidance in NUREG-1434, the licensee'has proposed to relocate all or portions of the following existing TS within-the improved TS:

1 Existina TS Title 3/4.4.1.1 Recirculation Loops 3/4.4.2.1 Safety / Relief Valves 3/4.4.2.2 Safety / Relief Valves Low-Low Set Function 3/4.4.7 Nain Steam Line Isolation Valves 3/4.4.9.1 Residual Heat Removal l

j 3/4.4.9.2 Cold Shutdown i

The more significant changes resulting from relocated items are as follows:

1.

The relief and low-low set valve function setpoints specify open and closed settings required to satisfy the valve safety function which is needed to prevent overpressurization of the nuclear steam system. The i

improved TS address these items within TS 3.3.6.5, " Relief and Low-Low

}

Set Instrumentation." The setpoint conditions for relief valve function l

of the S/RVs operability are consistent with the guidance provided in NUREG-1434 and are effectively retained within the improved TS.

2.

Existing SR 4.4.1.1.1 establishes operability requirements for recirculation system flow control valves through required testing that verifies valve stroke times and valve response to loss of control oil i

pressure. These requirements are moved to improved TS 3.4.2, " Flow Control Valves." The new LCO provides adequate requirements for these j

valves.

3.

Operability requirements specifying closing times for the main steam isolation valves and the limitations to operation with the valves inoperable are. relocated to improved TS 3.6.1.3

" Primary Containment i

Isolation Valves." The main steam isolation valve closure time limits j

are effectively retained in TS, and.this change is therefore acceptable.

4.

Existing TS 3/4.4.9.1 and 3/4.4.9.2 requirements to permit removing the 2

RHR shutdown cooling loop during hydrostatic testing is being relocated

]

to Note 3 of LCO 3.4.10, RHR Shutdown Cocling System - Cold Shutdown and LCO 3.10.1, Inservice Leak and Hydrostatic Testing Operation. The i

current requirements are effectively retained in the improved TS, and the change is therefore acceptable.

l j

1 i

72 i

The above changes are considered administrative changes in the location of the requirements within the improved.TS, and are therefore acceptable.

In accordance with the guidance in the Final Policy Statement, the licensee has proposed to relocate or reorganize all or pcrtions of the following j

existing TS to other licensee-controlled documents:

h-Existina TS Ilth f

f 3/4.4.1.1 Recirculation Loops

)

3/4.4.1.4 Idle Recirculation Loop Startup j

3/4.4.2.1 Safety / Relief. Valves l

3/4.4.2.2 Safety Relief Valves Low-Low Set Function i

3/4.4.3.1 Leakage Detection Systems j

3/4.4.3.2 Operational Leakage j

3/4.4.4 Chemistry 1

3/4.4.5 Specific Activity 3/4.4.6.1 Pressure / Temperature Limits 3/4.4.8 Structural Integrity 4

3/4.4.9.1 Hot Shutdown l

3/4.4.9.2 Cold Shutdown I

1.

The flow rate speed limits of an operating loop during idle loop startup j

j specified in existing TS 3.4.1.4 are not based on a required initial condition of any transient analysis. The loop flow rates do not reduce thermal stresses on vessel components; rather this item represents an l

operational limit that reduces the likelihood of a plant scram during idle loop startup. Avoiding operating regimes that could initiate plant protective systems does not directly relate to recirculation loop operability. These requirements have been relocated to the Bases.

Further, this change is made to be consistent with the NUREG-1434 l

presentation.

2.

Existing TS 3.4.2.1 Action b. requirements for stuck open S/RVs establish i

shutdown requirements if suppression pool temperatures approach the I

bounding assumptions of a Loss-of-Coolant Accident (LOCA).

If a stuck i

valve condition exists TS 3.4.2.1 Action b. requires the valve to be closed. The requirements are relocated to the Bases. The improved TS i

contain implicit requirements for remedial actions for suppression pool temperatures in improved TS LCO 3.6.2.1.

The LCO will require that the reactor Mode switch be placed in Shutdown if the temperature is greater than or equal to 110*F. The existing TS action is anticipatory to this.

i requirement in the event of a stud open S/RV, and preemptive in all cases. This action presents detailed methods of responding to an event 4

and is not a remedial action for failure to meet this LCO.

j 3.

Existing post-maintenance surveillance requirement 4.4.3.2.2.b has been relocated to the Bases. Any time the operability of a system or component has been affected by repair, maintenance or replacement, post-j maintenance testing is required by improved SR 3.0.1 and TS 3.4.6, RCS i

Pressure Isolation Valves (PIV) leakage, to demonstrate operability of the affected system or component. Therefore, changing to the explicit requirement for post-maintenance surveillance testing does not alter the i;

I-i

.1

73 I

improved TS requirements to establish operability before returning equipment to service.

4.

The offgas isotopic analyses specified by existing TS 3/4.4.5 Table 4.4.5-1 for xenon and krypton that are used to routinely monitor and i

trend coolant activity are relocated to plant procedures. The isotopic analyses required' for dose equivalent iodine-131 limits are retained in i

the specification because they are more controlling.. Separate limits for j

xenon and krypton do not ensure that the limits of the LCO are maintained. These requirements are relocated to the UFSAR or TS Bases.

Further, this change is made to be consistent with the NUREG-1434 presentation.

1 3

5.

Existing TS 4.4.6.1.1 monitors the reactor vessel pressure and metal temperature limits for the flange surfaces, bottom head outside surface -

j and bottom head inside surface as measured by the bottom head drain temperature. The limits specified by the composite PT-limits curve are contained in Figure 3.4.11-1 of the improved TS. The requirements being relocated to the procedures are details regarding the location of the i

sampling for performing surveillance testing and are~one method of 4

demonstrating the plant operating-limits for the reactor coolant pressure i

boundary; those limits are retained in the improved TS.

Further, this change is made to be consistent with the NUREG-1434 presentation.

l 6.

The reactor coolant system chemistry limits of existing TS 3/4.4.4 are relocated to the UFSAR. The reactor coolant chemistry program provides l

limits on particular chemical properties of the primary coolant, and surveillance practices to monitor those properties, to ensure that degradation of the reactor coolant pressure boundary is not exacerbated i

by poor chemistry conditions. However, degradation of the reactor i

coolant pressure boundary is a long-term process, and there are other, direct means to monitor and correct the degradation of the reactor coolant pressure boundary which are controlled by regulations and TS; for example, in-service inspection and primary coolant' leakage limits are 3

provided to prevent long-term degradation of the reactor coolant pressure i

boundary materials, and provide long term maintenance of acceptable j

structural conditions of the system. These limitations are not of immediate importance to the operator, and are not required to ensure i

operability of the reactor. coolant system pressure boundary.

Further, j

this change is made to be consistent with the NUREG-1434 presentation.

7.

The structural integrity inspections in existing TS.3.4.8 establish limiting conditions for operation to prevent long-term degradation of 4

ASME Code Class 1, 2, and 3 components. The inspection program associated with the TS requirements is performed on systems assumed to function to mitigate a design basis accident. However, the TS establish operability requirements for these same systems in addition to i

appropriate inspection requirements in improved TS 5.5.6 to ensure that i

structural degradation of safety systems will' be within limits. The l

specification limits in current TS.3.4.8 are not required to ensure operability of ASME Code class 1, 2, and 3 components. Therefore, the requirements specified in current TS 3.4.8 can be relocated the UFSAR, 4

i

.~.,-s.-w--

n,-

g--

y

,-~

.r

..,v..

.. E w. n,

, w m.mr, 4-,-#.-,,

e_-.--..,

lr i

e 74 i

i changes to which are controlled in accordance with 10 CFR 50.59.

i Further, this change is made to be consistent with the NUREG-1434 presentation.

8.

The drywell sump (equipment drain) monitoring system specified in existing TS 3.4.3.1 quantifies RCS identified leakage.

Relocation of.the details of the methods for monitoring systems do not impact RCS-i-

operability. Therefore, the requirement for the drywell equipment drain

]

sump monitoring system has been relocated to the UFSAR and TS Bases l

4 support of compliance with the limits for RCS Operational' Leakage in.

improved TS 3.4.5.

I 9.

The requirement of-existing TS 3.4.1.1 to suspend any power or. loop flow-l increases when the single loop operating temperature differences limits i

of SR 4.4.1.1.5 are exceeded is relocated to UFSAR or TS Bases.

In.

addition, the. time allowed to restore temperatures to within limits in j

low power and low flow conditions is revised to be consistent with other i

operational conditions by revising the requirement to " suspend the 1

THERMAL POWER or recirculation loop flow increase" (with no required j

completion time) to " restore parameters to within limits within 30 minutes." The need for both required actions are to prevent undue j

thermal stresses on the RPV and the spray nozzles.

Improved TS 3.4.1, j

3.4.2 and 3.4.11 specify TS limits to ensure thermal stresses on the RPV-j and flow nozzles are within those values accepted by the staff for analyzed accidents and transients.

s i

i

10. Existing TS 3.4.1.1 requirements specify operator actions to insert i

control rods or increase flow by opening flow control valves if j

j recircuiation pumps are operating on fast speed. These detailed methods

)

i for performing the required action to restore power and flow limits to within the acceptable region of the specified figure are relocated out of i

improved TS 3.4.1 to'UFSAR and TS Bases. Although the control of this i

action by procedure may provide some operational. flexibility, relocation 4

to UFSAR and TS Bases is acceptable because sufficient control over

]

important limits are maintained by the LCO.

1 l

11. Existing TS 3.4.1.1.b volumetric flow rate limits and the requirements in t

Action f. to reduce loop volumetric flow rate have been relocattd to j

UFSAR and TS Bases. The additional flow rate limits restrict rtactor.

vessel internal vibrations to within acceptable limits during coMitions of single loop operation. Total core flow as a function of thermal power j:

limits are established in improved TS figure 3.4.1-1.

The additional relocated limits are not evaluated in the transient analyses for single 1

loop operation for demonstrating sufficient flow to maintain fuel thermal margins during transients provided MCPR limits are modified. Since MCPR limits are required to be modified by improved TS 3.4.1 relocation of

^

these items is acceptable.

Existing TS limits and restrictions for i

recirculation loop flow, as discussed in items 9,10, and 11 above, do j

not directly relate to recirculation loop operability. Therefore, these requirements have been relocated to the TS Bases. Changes, as necessary, to the TS Bases are made pursuant to 10 CFR 50.59.

i i

e 1

i i

l 75 j'

l

12. A note to existing TS 3.4.9.2 requires at least one of two required j

shutdown cooling loops to have an associated operable diesel generator.

This requirement was added to provide for additional capability for mitigation of a loss of decay heat removal event. However, unlimited.

j continued operation is allowed following the loss of RHR shutdown cooling using any alternate decay heat removal method. During ' shutdown, TS allow various required equipment to be powered by either division of class IE power or split between the divisions without regard for Wich diesel j

generator is operable. Therefore, the requirements spet'.? Nd in current TS 3/4.4.9.2 for onsite AC power are not. required to suppo:

he i

operability of a TS system assumed to function during an acJ oent.

l Therefore, the policy statement criterion for inclusion in TS is not met i

and these requirements can therefore be relocated to TS Bases, changes j

for which are controlled in accordance with 10 CFR 50.59.

1

13. 1.ists of valves specified as pressure isolation valves in existing TS l

3.4.3.2 are details relating to system design and purpose have been J

relocated to the Bases or procedures. The design features and system i

operation associated with these valves are also described in the UFSAR.

Changes to the Bases will be controlled by the provisions of the proposed l-Bases Control Process in Chapter 5 of the. Technical Specifications.

14. Requirements in existing TS 3.4.3.2 specifying leak testing pursuant to ASME Section XI for pressure isolation valves are required by improved TS SR 3.4.6.1.

Details of the interval for performing this surveillance are I

relocated to the Inservice Testing (IST) Program.. Additionally, the IST

- Program will be controlled by the provisions of the proposed l

Administrative Controls in Chapter 5 of the Technical Specifications.

15. Operability requirements for leakage pressure monitors on high/ low pressure interface valves are deleted from existing TS 4.4.3.2.3, Table 1

i 3.4.3.2-2.

The alarm instrumentation does not relate directly to operability of the system isolation valves.

Improved TS 3.4.6 provides sufficient control over important valve operability requirements to ensure leakage from each reactor coolant system pressure isolation valve j

is within limits. Control of the operability and the required actions for monitoring and indication instruments, and test equipment are j

therefore adequately controlled by the UFSAR and TS Bases, j

The above relocated requirements relating to the reactor coolant system are 4

i not required to be in the TS under 10 CFR 50.36, and are not required to i

obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety. Further, they do not fall within any of the four criteria set forth in the Commission's Final Policy j

Statement, discussed in the Introduction above.

In addition, the staff finds i

that sufficient regulatory controls exist under 10 CFR 50.59. Accordingly, the staff has concluded that these requirements may be relocated from the TS l

to the licensee's TS Bases or to the UFSAR,'as' applicable.

l i

I

- ~

76 C..

More Restrictive Requirements By electing to implement the NUREG-1434 Section 3.4 Specifications, the-licensee has adopted a number of more restrictive conditions than are required by the existing TS. The more significant conditions are the following:

1.

A specific completion time for the engineering evaluation required by existing TS 3/4.4.6.1 actions is proposed. The proposed time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is considered reasonable for operation in Modes 1, 2 and 3 because the limits represent controls on long term vessel fatigue and usage factors.

In Modes 4 and 5, the proposed frequency would prevent entry in the operating modes, which is consistent with the current LCO 3.0.4.

2.

The proposed change modifies Item 2 of existing TS Table 4.4.5-1 to change the frequency for isotopic analysis for dose eqWvalent I-131 concentration limits from at least once per 31 days to at least once per seven days. -This modification compensates for deleting the improved TS 3.4.8 requirement to verify every seven days that gross specific activity remains less than or equal to 100/E-bar microcuries per gram thereby ensuring offsite ' doses will remain within a small fraction of the limits of 10 CFR Part 100.

3.

A Note to the Applicability requirements of existing TS 3/4.4.6.2 indicates that the reactor steam dome pressure limit is not applicable during anticipated transients. Although design basis accidents are not assumed to occur during an anticipated transient, the reactor steam dome pressure limit is applicable during such transients, and the required actions should be taken to improve mitigation of the transient.

Therefore, improved TS 3.4.12 does not-permit this exception to the steam dome pressure limits.

The staff has reviewed the more restrictive requirements and concludes that they result in an enhancement to the improved TS. Therefore, the more restrictive requirements are acceptable.

D.

Less Restrictive Requirements i

The licensee in electing to implement the NUREG-1434 Section 3.4 Specifications has adopted a number of less restrictive conditions than'are allowed by the existing TS. The more significant conditions are the i

following:

l.

A Note is added to improved TS 3.4.1 that proposes to increase the time allowed to complete the required limit and setpoint modifications after a change from two loop operation to one loop operation. The current eight hours can be insufficient to complete the required modifications and does not allow for orderly briefings, planning, and discussion among the involved plant organizations that is needed when change-over to single loop operations is required by unforeseen circumstances. The proposed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is the length of time allowed by improved TS SR 3.2.1.1 to verify the average planar linear heat generation rate and improved TS-SR 3.2.2.1 to verify MCPR parameters are within limits when initially entering their

4 4

77 applicable conditions. During the proposed allowed out-of-service time for operating in unverified conditions or conditions outside the limits there is an insignificant impact on safety, therefore the staff concludes the proposed change is acceptable.

2.

A Mode 3 exception for the operational leakage requirements of existing TS 3/4.4.3.2 for PIVs is included for valves in the shutdown cooling flow path when needed for the shutdown cooling function.

This change resolves a conflict in the current specifications that requires shutdown cooling flow path isolation if the pressure isolation valve leakage is not within i

limits, even with reactor coolant system pressure below the RHR cut-in permissive pressure when shutdown cooling is required to be operable and operating. Although alternative methods of decay heat removal could be established, shutdown cooling is the preferred method.

Further,_its use with leaky pressure isolation valves poses no risk at low pressure since the high to low pressure interface condition does not exist.

t 3.

Applicability requirements in the improved TS for primary coolant specific activity are limited to those conditions which represent a potential for release of significant quantities of radioactive coolant to the environment. Mode 4 is deleted from the existing TS 3.4.5 during conditions that the reactor is not pressurized and for which the potential for leakage is significantly reduced.

In Modes 2 and 3, with the main steam lines isolated, no escape path exists for significant releases and requirements for limiting the specific activity are not required. The improved TS required actions are therefore also modified to reflect these applicability changes, and an option for exiting the applicable Modes is provided for cases where isolation is not desired.

4.

The improved TS delete existing TS LCO 3.4.5.b associated actions and surveillance requirements, which require gross specific activity for non-l iodines in the reactor coolant to be limited to less than or equal to 100/E-bar pCi/ gram. This proposed change also deletes Item 1 of Table 4.4.5-1 requiring gross beta and gamma activity determination at least once'per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The Bases for existing TS 3.4.5 state that the intent of the requirement to limit the specific activity of the reactor coolant is to ensure that whole body and thyroid doses at the site boundary would not exceed a small fraction of the limits stated in 10 CFR Part 100 (i.e.,10% of 25 rem and 300 rem, respectively) in the event of a main steam line failure outside containment. To ensure that offsite thyroid doses do not exceed 30 rem, reactor coolant Dose Equivalent Iodine-131 (DEI) is limited to j

less than or equal to 0.2 Ci/ gram. Likewise, reactor coolant gross specific activity is limited to less than or equal to 100/E-bar Ci/ gram to ensure that whole body doses do not exceed 2.5 rem.

LCO 3.7.8.2 (ITS LC0 3.7.5) associated with radioactive effluents requires that the gross gamma radioactivity rate of the noble gases measured prior to the holdup pipe be limited to less than or equal to 289 millicuries /second. The Bases for LCO 3.7.8.2 state that restricting the gross radioactivity rate of noble gases from the main condenser provides i

a

~

l' 9

c 78 j

reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of 10 CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment.

j The offgas treatment system, as required by improved TS 3.7.5 ensures that the reactor coolant gross specific activity is maintained at a 1

sufficiently low level to preclude.offsite doses from exceeding a small fraction of the limits of 10 CFR Part 100 in the event of a main-steam i

line failure. Therefore, existing TS 3.4.5.b is redundant and elimination of this TS is acceptable. Additional assurance that the

.i offsite. doses will not exceed a small fraction of the 10 CFR Part 100 limits is provided by increasing the frequency of sampling and analysis of the reactor coolant for DEI from at least once per 31 days to-at least once per seven days.

5.

A Note is added to the required actions for Condition A of improved TS 3.4.8 to indicate that LCO 3.0.4 is not applicable. Restrictions on entry into TS applicable Modes are not required given that the expected i

response to the. condition of not meeting the dose equivalent iodine or gross specific activity limits is restoration of LCO compliance within 1

the allowed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, a reasonable time for temporary coolant activity increases (iodine spikes or crud bursts) to be cleaned up by the normal processing systems. Further, since the LCO limits assure the dose due to a LOCA would be a small fraction of the 10 CFR Part 100 limit, continued operation with the limits of the specification not met during the 12 1

hours allowed to restore the limits should not have represent a significant impact on the health and safety of the public; if the limits are not met at the end of the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the main steam lines must be isolated or the reactor must be shutdown.

6.

Existing TS 3/4.4.3.2 requires at least two closed manual, or deactivated automatic valves for the isolation of the high pressure to low pressure interface portions of the RCS. Once the initial isolation of the flow path is complete, a specified period of time is allowed in improved TS 3.4.6 to complete the isolation by use of a second device. The additional time is provided based on the time required to complete the required action, the low probability of a second valve failing during this time period, and the low probability of a pressure boundary rupture of the low pressure emergency core cooling system (ECCS) piping when pressurized to reactor pressure.

7.

The existing TS 3/4.4.6.2 limit for reactor steam dome pressure is changed from less than 1045 psig to include a pressure equivalent to 1045 psig in improved TS 3.4.12.

This is an inconsequential change that is considered less restrictive since technically it increases the range of the allowable pressure. However, the change is consistent with the safety analysis assumptions of reactor steam dome pressure at 1045 psig, therefore the allowed increase in pressure is acceptable.

8.

Changing from reactor power operation conditions to shutdown requires reducing reactor coolant temperature using the RHR shutdown cooling i

1 79 i:

system once reactor pressure is reduced below the RHR low pressure permissive setpoint.

In general, TS LCO 3.0.4 and SR 3.0.4 requirements prohibit mode changes unless the applicable conditions and SRs for the t

LCO are met before a mode change is made.

Improved TS SR 3.4.9.1 includes a note that the SR is not required to be met until two hours after reactor pressure is less than the RHR pressure permissive.

Therefore, the notation, which permits entry into the conditions of i

improved TS 3.4.9 while depending on the actions because SR 3.4.9.1 is l

not met, provides the time operators need to place the RHR shutdown i

j cooling system in operation following the reduction of pressure to below l

the low pressure permissive cut-in setpoint. The additional two hours to operate without the LC0 being met is consistent with the requirements retained in the improved TS for the condition of one or two RHR shutdown cooling subsystems inoperable.

l j

i j

9.

Existing TS 4.4.1.2.c required testing of jet pump distribution using

)

diffuser to lower plenum differential pressure is verified to be within l

10% of established patterns is proposed to be modified in improved TS SR

)

3.4. 3.1.c to allow a second established limit to be met for the operability determination. This second limit is based on a flow

]

comparison. Because the second proposed method.also establishes jet pump i

structural integrity thereby ensuring the ability to flood the reactor i

vessel to two-thirds of core height, the staff concludes the proposed change is acceptable.

l

10. The time requirements in existing TS 3.4.1.1 for restoring the limits and setpoints modified due to single loop operation following a return to two l

loop operation are deleted from improved TS 3.4.1.

The existing TS 4

establish single loop operation limits and setpoints that are conservative with respect to the two loop operation limits and setpoints.

i Operation is acceptable with the more conservative limits for an j

unlimited time. Therefore, the constraint to establish limits and j

setpoints within a specified time upon restoration of two loop operation is unnecessary.

S i

11. The existing TS 3.4.1.1 time requirement to verify temperature "within every hour" during thermal power or recirculation loop flow increases is deleted from improved TS 3.4.1.

The improved TS establish frequencies for SRs that are based on the reasonable expectation that periodic retesting will generally confirm operability or compliance with the limits. Therefore, the requirement to perform reverification of required 4

setpoint and limit adjustments during changes in reactor conditions once i

setpoints and limits are established is unnecessary.

i 12.

Existing TS 4.4.1.3 requires the recirculation loop flow mismatch to be j

i within required limits. This requirement is changed by a note to L

improved TS SR 3.4.1.1 to allow 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for performance of the test c

after the applicable mode is entered because this surveillance cannot be

]

performed prior to its applicability. This frequency is consistent with the frequency for jet pump operability verification and has been shown by

{'

operating experience to be adequate to detect off normal jet pump loop flows in a timely manner. Therefore, an allowance for time to initiate 1

i a

G d

t

]

80 l

the frequency is required to avoid intentional entry into the LCO j.

Condition each time the second recirculation pump is started.is acceptable.

13. The Action and SR are. changed in improved TS 3.4.10 to agree with the existing TS LCO 3.4.9.2 requirements. The LCO specifically includes the i

operation of a recirculation pump as an acceptable method for assuring-the necessary flow conditions. However,-this.is not currently recognized by either Actions a. or b. or existing TS SR 4.4.9.2.

l 1

1 The above less restrictive requirements have been reviewed by the staff and j

' have been found to be acceptable, because they do not present a significant i

safety question in the operation of the plant. The TS requirements that i

remain are consistent with current licensing practices, operating experience t

and plant accident and transient analyses, and provide reasonable assurance that the public health and safety will be protected.

j j

3.5 Emergency Core Cooling Systems i

f A.

Significant Administrative Changes j

In accordance with the guidance in the Final Policy Statement, the licensee.

I j

has proposed administrative changes to the existing technical specifications (TS) to bring them into conformance with the improved TS. These changes are l

4 as follows e

i 1.

The footnote associated with the action statements of existing TS 3.5.1 i

provides alternate actions when two or more residual heat removal l

subsystems are inoperable and cold shutdown cannot be attained. This i

footnote is removed in the improved TS since it provides unnecessary i

i duplication of existing TS 3.4.9.1 actions and the improved TS 3.4.9-i actions. Also, it contains no additional restrictions on the operation i

of the plant, and could be interpreted as a relaxation of the j

requirements to achieve cold shutdown. Since sufficient direction is j

provided elsewhere in the improved TS, this potential interpretation is deleted and this administrative change is found acceptable.

L

)

2.

Improved TS 3.5.1 Condition H provides direction for various 1

interrelationships between the Division 1 and/or 2 emergency core cooling i

l system (ECCS) subsystems and the Division 3 system. The action requires j

entry into Limiting Condition for Operation (LCO) 3.0.3 for various combinations of inoperable components which are consistent with the 4

present actions for the same combinations.

i 4

3.

Existing TS 4.5.3.2 requires that cartain surveillances be performed on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> frequency when in Modes 4,r 5 and the suppression pool water level is less than the specified simit. The incorporation of existing TS 4.5.3.2 into improved TS LCO 3.5.2 and SR 3.5.2.2 represents a change in 2

the surveillance frequency. However, some of the surveillance details will be required more frequently in that verification of existing TS LCO l

3.5.3 items b.1, b.2 and footnote

• will be required continuously as part j

I 81 l

of the improved TS 3.5.3 Applicability, and the remaining details of High Pressure Core Spray (HPCS) operability will continue to be verified every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Therefore, the new requirements are only in a different format l

from the current specifications with no change in. intent or effect.

j Thus, this administrative change is found to be acceptable.

4.

Existing TS 3.5.1 Action h requires that in the event an ECCS system is' 4

actuated and injects water into the reactor coolant system, a Special Report be prepared and submitted to the Commission pursuant to 3

Specification 6.9.2 within 90 days. The requirement for a Special Report 3

i following an ECCS injection, the associated details of the report, and i

when it should be submitted can be adequately controlled by the s

licensee's administrative controls.

10 CFR 50.73 already provides the

)

requirement for the licensee to submit a Licensee Event Report in the j

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. Removing a duplicate requirement from the TS has no t

impact on assuring safe operation of the facility since the requirement to submit a report to the Commission still exists in 10 CFR 50.73.

l Therefore, the staff concludes that.these regulatory requirements provide sufficient control of these provisions and removing them from the TS is acceptable.

5.

The requirement of existing TS 3.5.2 Action b and 3.5.3 Action b to j

establish containment integrity during fuel handling would appear to.

i.

provide a period of time (eight hours) during which integrity could be j

violated even if capable of being maintained. Additionally 'if the plant status is such that integrity is not capable of being established within 4

)

eight hours, the existing TS action results in non-compliance with the TS and a requirement for a licensee event report. The intent of the action i

is believed to be more appropriately presented in NUREG-1434 TS 3.5.2 i

required actions. These actions impose a significantly more conservative requirement to establish and maintain the. containment boundary. No j

longer would the provision permitting a licensee to violate the boundary i

for up to eight hours appear to exist. The licensee has modified i

improved TS 3.5.2 Required Action D to account for the plant specific I

containment design and the required boundary for this condition as defined by the current licensing basis.

1 6.

Just as during power operations, some unisolated penetrations may be administrative 1y open with capability for closure. These isolation valves would not have instrumentation required to be operable. Other i

penetrations may utilize isolation devices other than valves. To accommodate the GGNS design plant-specific wording is used in LCO 3.5.2, j

Required Action D 3 and Bases. Not all penetrations are assumed to be isolated or isolatable; only those that are required (i.e., assumed to perform an isolation function) are necessary. Appropriate plant-specific wording has been provided. The staff has reviewed these changes and J

finds that they conform to the current licensing basis of the plant.

Since the current licensing basis has previously been found acceptable from a safety standpoint, the staff finds the proposed changes j

acceptable.

i i

i

,.,m

.,m,

+

4--,...,- e.v e-

~,

i*

l 82 i

7.

Improved TS ras allow one-hour to verify RCIC-or HPCS operability. This j

time does not exceed the time that could be inferred from the existing TS for concurrent HPCS and RCIC inoperability which requires.the plant to i

4 begin following LCO 3.0.3.

LCO 3.0.3 allows an hour before the licensee i

must begin shutting the plant down. This period is consistent with the completion times for ras in improved TS'LCO 3.3.5.1, LCO 3.6.3.1 RA B.1, LC0 3.6.3.3 RA B.1, and LCO 3.8.1 RA B.2.

These LCOs allow the licensee 4

at least an hour to verify the operability of multiple systems even l

though it may immediately know this:information.

Since the above changes result in the same limits as the current requirements, or they represent an enhanced presentation of the existing TS intent, the i

improved TS changes are purely administrative and are acceptable.

B.

Relocated Requirements i

In accordance with the guidance in NUREG-1434, the licensee proposed to relocate all or portions of the following existing TS within' the improved TS.

4 f

Existina TS Title 3/4.5.1 ECCS - Operating 3/4.5.3 Suppression Pool The more significant changes resulting from relocated items are as follows:

j 1.

Existing TS 4.5.1.a.3 includes requirements to periodically verify that i

each valve (manual, power operated, or automatic) in the Low Pressure Core Spray (LPCS), Low Pressure Coolant Injection (LPCI) and HPCS flow path that is not locked, sealed, or otherwise secured in position, is in its correct position. The licensca has proposed moving this surveillance requirement to improved TS SR.',.5.1.2 and SR 3.5.2.4.

2.

Existing TS 4.5.1.c.4 specifies that the injection valves for LPCI and low-pressure core spray (LPCS) be tested in accordance with existing TS i

4.0.5.

Improved TS 5.5.6, " Inservice Testing Program," addresses the j

4 requirements for existing TS 4.0.5; thus, existing TS injection valve j

testing requirements are included in the improved TS.

The above changes are considered purely administrative changes in the location of the requirements in the improved TS and are therefore, acceptable.

}

In accordance with the guidance in the Final. Policy Statement, the licensee j

has proposed to relocate or reorganize all or portions of the following j

existing TS to other licensee-controlled documents:

Existina TS Title 3/4.5.1 ECCS - Operating 3/4 5.2 ECCS - Shutdown 3/4.5.3 Suppression Pool 1

i 7

. -. ~

, -, -., _, ~,. _ -,.. -,

l i

j

'83 The more significant changes resulting from relocated items are as follows:

l

)!

1.

The ECCS " keep-filled" pressure instrumentation, ECCS delta-pressure instrumentation and the automatic depressurization system (ADS)

{

accumulator low pressure alarm system instrumentation specified in existing TS 3.5.1 Actions f, g and i and the associated surveillances in existing TS 4.5.1 do not relate directly to the respective. system operability.

In general NUREG-1434 does not specify indication-only or test equipment to be operable to support operability of a system or

.I 8

com,)onent.

In this case, if the ADS accumulator low pressure alare system instrumentation becomes inoperable, alternative means are i

available to the operators to verify accumulator pressure and hence, i

system operability. Control of the availability of, and necessary I

compensatory activitics if not available (such as more frequent venting of the system), for indications, monitoring instruments, alarms, and test i

equipment are addressed by the UFSAR and TS Bases. Therefore, J

i operability requirements for this instrumentation, along with the supporting actions and surveillances are relocated to the UFSAR and TS Bases. Further, the change is made to be consistent with the.NUREG-1434 presentation.

j 2.

Existing TS 3.5.3.b.2 and 3.5.3 Action b specify that the reactor mode switch shall be locked in the shutdown or refuel position for Modes 4 or 5 operation. Reactor mode switch operability.is included as part of the operability of the associated interlocks required by improved TS 3.9.1 i

and 3.9.2.

Movement of the reactor mode switch from the shutdown position is adequately controlled by the improved TS Modes definition t

Table 1.1-1.

The requirement to " lock" the mode switch can be relocated j

to the UFSAR or TS Bases.

i l

3.

Existing TS 3.5.1,. 3.5.2, 3.5.3.b.4 and 3.7.3 specify the details I

relating to HPCS, LPCS, LPCI and RCIC system design and purpose.

Existing TS 4.5.1.a, 4.5.1.c, 4.5.1.d, 4.5.1.e 4.7.3.c specify the methods for performing these surveillances. The methods for determining i

ECCS system operability depend on plant design features and system j

operation. The design features and system operation which dictate the methods in the existing TS are described in the UFSAR and improved TS Bases.

Since the procedural details of how a specific surveillance is-performed are not located in the-improved TS, the methods, system design, and purpose can be relocated to the UFSAR and TS Bases, and the details on how to accomplish the surveillance can be relocated out of the TS.

Thus, the details of the methods for performing the above surveillances j

can be relocated to the UFSAR and TS Bases.

4.

Existing TS 4.5.1.d.3 specifies air pressure decay test on the ADS. This i

SR has been relocated to plant procedures since the requirement does not directly relate to ADS operability. The system may be able to perform i

its required safety function in the short term (i.e., less than 7 days),

yet would be required to be declared administrative 1y inoperable due to the existing TS-requirement.

Improved TS SR 3.5.1.3 addresses whether j

sufficient air pressure is available to permit the ADS valves to actuate'

{

if an accident occurs. Although the surveillance being relocated is not t

.s i

l 84

[

a requirement for operability, it will continue to enable the licensee to verify the capability of the ADS air system to retain pressure. Thus, i

the requirement can be adequately controlled outside of TS.

i l

5.

The suppression pool' water level instruments stated in existing TS 3.5.3 Actions e and d, and existing TS 4.5.3.1.b and 4.5.3.1.c may not relate directly to the operability of their respective systems. The improved TS do not specify indication-only or test equipment to be operable to i

support operability of a system or component.

Plant operational procedures address the control of the availability of indications, i

monitoring instruments,' alarms, and test equipment.- The procedures also address necessary compensatory activities if these indications and i

instruments are not available. Therefore, these instruments and the supporting surveillances and actions may be relocated from the improved TS to the UFSAR.

6.

Existing TS 3.5.2, 3.5.2 Action a, and 3.5.2 footnote ** require one-of j

two operable ECCS subsystems or systems to have an associated operable

~

diesel generator and require in RA a. that the remaining ECCS subsystems j

and systems to be capable of automatic initiation.. This requirement was-added to the GGNS TS to provide for additional capability to mitigate a 1

i.

drain down event while the alternate decay heat removal system (ADHRS) is l.

operating. However, unlimited continued operation has also been allowed i

i after one or both of the required ECCS subsystems or: systems are lost if i

the licensee suspends operations that could drain the reactor vessel or l

]

if secondary containment integrity is established, as applicable.

Therefore, the change to allow ADHRS to be used did not reduce the j

capability to mitigate a drain-down event that results from either (1) a loss of an ECCS from the loss of off-site power for the allowed i

conditions controlling unlimited continued operation or (2) a loss of the 1

i automatic initiation capability.

Most outage activities are planned so that the operable equipment during a shutdown is powered by the one required operable diesel generator, but j

such planning is not required. The STS allow various required pieces of j

equipment to be powered by either division or split between the divisions j

without regard for which diesel generator is operable. The STS are lenient in this matter because during an outage, the reactor coolant pressure boundary contains significantly less energy, the operating temperature and pressure of the reactor coolant are very low, and the corresponding stresses are minimal. Thus, the loss of ECCS would result in minimal consequences.

1 Although NUREG-1434 does not require that the one operable ECCS system be powered by an operable associated diesel generator, The licensee maintains this equipment arrangement during shutdown conditions (Modes 4 and 5). However, this commitment can be administrative 1y maintained as is currently being done at other plants. Therefore the control of this i

commitment under 10 CFR 50.59 is acceptable. The commitment can be p

relocated to the TS Bases.

i.

i

__m_

j 85 7.

Existing TS 3.5.2 and 3.5.3 restrict operation to prevent the licensee from using LCO 3.0.4 to change modes.

Improved TS 3.5.2 will control such changes administrative 1y, and therefore the requirement is removed l

from TS. LCO 3.0.4 does not restrict mode changes for. conditions of the i

actions during which the improved TS allow unlimited continued operation l

in the applicable modes. However, such. mode changes are not prudent and would not be allowed by the proposed administrative controls. Good j

practice in accordance with administrative controls will dictate when plant startup and mode changes are allowed with inoperable equipment.

This practice is consistent with the NRC position stated in Generic Letter 87-09.

j The above relocated requirements relating to the ECCS systems are not required to be in the TS under 10 CFR 50.36, and are not required to obviate the.

l possibility of an abnormal situation or event giving rise to an immediate' threat to the public health and safety. Further, they do not fall within any i

of the four criteria set forth in the Comission's Final Policy Statement, i

discussed in the Introduction above.

In addition, the staff finds'that l

sufficient regulatory controls exist under TS 5.5.11 and 10 CFR 50.59.

l.

Accordingly,.the staff has concluded that these requirements may be relocated l

from the TS to the licensee's TS Bases, UFSAR or plant procedures, as j

applicable.

4 C.

More Restrictive Requirements.

By electing to implement the NUREG-1434 Section 3.5 Specifications, the j

licensee has adopted a number of more restrictive conditions than are required l

by the existing TS. The more significant conditions are the following:

5 i

1.

Improved TS SR 3.5.1.3 has been added to verify that ADS air accumulator supply pressure is greater than or equal to 150 psig. This is a new i

surveillance requirement which verifies that sufficient air pressure i

exists in the ADS accumulator supply for reliable operation of ADS.

2 Since this is a new surveillance requirement, it is an added restriction i

to plant operations.

i l

2.

The frequency for existing TS 4.5.1.d.2 in the improved TS has been i

changed to be performed on a staggered test basis for each valve j

solenoid. Currently the test could be done each time with the same i

solenoid. The proposed test assures that each valve solenoid will be used alternately every 18 months to cycle the valve.

This limits the i

number of times the valves must be cycled, and thereby limits potential i

damage to the valves.

I i

The staff has reviewed these more restrictive requirements and concludes that i

they result in an enhancement to the existing TS. - Therefore, these more i

restrictive requirements are acceptable.

1 I

L.

i l

86 j

D.

Less Restrictive Requirements The licensee in electing to implement NUREG-1434, Section 3.5 Specifications has proposed a number of less restrictive conditions than are allowed by the j

existing TS. The more significant conditions are the following:

I 1.

Two new actions are being added to existing TS 3.5.1: (1) for the condition of one ADS valve inoperable coincident with one low pressure ECCS injection / spray subsystem, and (2) for the condition of HPCS inoperable coincident with one LPCI subsystem. The analysis summarized in UFSAR Section 6.3.3 demonstrates that adequate core cooling is provided by the operable HPCS or ADS system and the remaining operable low pressure injection / spray systems.

However the redundancy has been reduced such that another single failure may not maintain the ability to provide adequate core cooling. Existing TS 3.5.1 recognizes the j

j diversity of the ECCS and permits continued plant operation with a similar loss of redundancy. Specifically, existing TS 3.5.1 permits.

i continued plant operation for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with two ECCS low pressure injection / spray systems inoperable. The proposed actions permit a-limited amount of time for equipment trouble-shooting and repairs as l

opposed to requiring an immediate plant shutdown in accordance with TS j

3.0.3.

Due to the diversity and redundancy of the ECCS, these actions

]

are considered reasonable in lieu of the risks associated with a forced plant shutdown. Therefore, the same allowable outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> i

has been assigned to restore either the inoperable ADS valve or the inoperable low pressure injection / spray system to operability.

I 2.

The relaxations specified in existing TS 3.5.2 Applicability footnote

• j and existing TS 3.5.3 Applicability footnote

• that the ECCS and suppression pool respectively are not required to be operable if water t

I level is maintained within the limits of existing TS 3.9.9 are being deleted. When in Modes 4 and 5 the suppression pool only provides i

support for operability of the ECCS. As described in UFSAR Section 9.1.3.3, the ECCS systems are not required for either normal or emergency 3

j makeup to the spent fuel pools. Thus, the apparent requirement linking j

ECCS and suppression pool operability with spent fuel pool water level is unnecessary.

a 3.

Existing TS 3.5.2 Action b requires that Core Alterations be suspended if j

both of the required ECCS subsystems / systems are inoperable. The i

licensee has proposed deleting the requirement that Core Alterations be suspended.

Existing TS 3.5.2 does not require the ECCS to be operable if l

the reactor vessel head is removed, the cavity is flooded, the reactor cavity to steam dryer pool gate is opened and water level is maintained at least 23 feet over the top of the reactor pressure vessel flange.

j Therefore, under these conditions, ECCS is not required to be operable and the proposed change is acceptable. Beyond these specific conditions, improved TS 3.9 LCOs provide requirements to ensure safe operation during l

Core Alterations including required water level above the reactor i

pressure vessel flange.

Specifically, improved TS 3.9 LCOs recognize the need for plant operation without operation of the ECCS and permits j

continued refueling activities provided that alternative methods of decay

)

f.

c t

r 4

f

]

87 1-heat removal are available. Therefore, adequate' precautions are being provided to ensure core cooling and the requirement to suspend Core 3

Alterations may be deleted.

)i 4.

Existing TS 3.5.3 Action b requires that Core Alterations be suspended if the suppression pool water level falls below specified limits. The i

licensee has proposed deleting the requirement that Core Alterations be j

suspended.

Existing TS 3.5.3 does not require the suppression pool to be operable if the reactor vessel head is removed,'the cavity is flooded,

)

e the reactor cavity to steam dryer pool gate is opened and water level is maintained at least 23 feet over the top of the reactor pressure vessel i

2 flange. Therefore, under these conditions, the suppression pool is not required to be operable and the proposed change is acceptable. As described above, improved TS 3.9 LCOs provide requirements to ensure safe i

operation during Core Alterations including required water level above the reactor pressure vessel flange..Therefore, the requirement to

{

suspend Core Alterations may be deleted.

l 5.

A Note clarifying the alignment requirements of the LPCI subsystems has been included in improved TS SR 3.5.1.2 (ECCS Operating) and SR 3.5.2.4 i

(ECCS Shutdown). The Note to improved TS SR 3.5.1.2 is similar to l

existing TS 3.5.2 Applicability footnote # in that the proposed Note allows LPCI subsystems to be considered. Operable during alignment and operation for decay heat removal with reactor steam done pressure less jl than the residual heat removal cut in permissive pressure in Mode 3, if capable of being manually realigned and not otherwise inoperable.

j Similarly, the Note to improved SR 3.5.2.4 allows operation of a single j

LPCI subsystem in operation for decay heat removal during Modes 4 and 5.

Because manual valve positioning removes the capability of the subsystems s

to respond automatically, the subsystems would be considered inoperable l

without this Note. Although no specific analysis of this condition has been performed, the allowance provided by the Note is acceptable because

{

the return to operability entails only the repositioning of valves, either remote or locally, and the energy requiring dissipation in Modes i

3, 4, and 5 below 150 psig, is considerably less than that at 100% power l

with normal operating temperature and pressure.

3 6.

The pressure specified in existing TS 3.5.1 at which ADS is required to I

be operable and in existing TS 3.7.3 at which RCIC is required to be L

operable is increased to 150 psig to provide consistency of the j

operability requirements for all ECCS and reactor core isolation cooling L

system equipment. Small break loss of coolant accidents at low pressures (i.e., between 100 and 150 psig) are bounded by analyses performed at higher pressures. The ADS is required to operate to lower the pressure i

sufficiently so that the LPCI and low pressure core spray systems can provide makeup to mitigate such accidents.. Since these systems can begin to inject water into the reactor pressure vessel at pressures well above

]

150 psig, there is no safety significance in the ADS not being operable between 100 and 150 psig.

The above less restrictive requirements have been reviewed by the staff and j

have been found to be acceptable, because they do not present a significant i

i

1 1.

i 88 i.

safety question in the operation of the plant. The TS requirements that' l

remain are consistent with current licensing practices, operating experience 4

and plant accident and transient analyses, and provide reasonable assurance that the public health and safety will be protected.

3.6 Containment Systems q

A.

Significant Administrative Changes i

In accordance with the guidance in the Final Policy Statement, the licensee

)

has proposed administrative changes to the existing technical specifications i

(TS) to bring them into conformance with the improved TS. These changes are as follows:

i 1.

The existing TS 3/4.6 Specifications contain details which are also found in Appendix J to 10 CFR Part 50. The regulations require licensee compliance and cannot be revised by the licensee. The improved TS state that requirements are "... in accordance with Appendix J and approved j

exemptions." Therefore, direct reference to Appendix J eliminates the l

need for repetitious and unnecessary details within the improved TS.

This.is consistent with the guidance of NUREG-1434 on presentation and is j

administrative in nature.

2.

The deletion of the Section 1.0 definitions for Primary Containment l

Integrity, Secondary Containment Integrity, and Drywell Integrity have j

been replaced by separate Limiting Condition for Operation (LCO) i requirements that the Primary Containment (LCO 3.6.1.1), Secondary Containment (LCO 3.6.4.1), and Drywell (LCO 3.6.5.1), respectively, shall 1

be operable in the applicable Modes.

In addition, the existing TS for i

Primary Containment and Drywell Structural Integrity have been i

4 incorporated as equivalent surveillance requirements within each of the

]

respective improved TS LCOs noted above. The previous elements of each of the " Integrity" definitions are now individually specified in their separate improved TS LCOs, such as air locks (LCOs 3.6.1.2 and 3.6.5.2);

{,

isolation valves (LCOs 3.6.1.3, 3.6.4.2, and 3.6.5.3); suppression pool (LCO 3.6.2.1 and 3.6.2.2); etc. The operability requirements collectively defined in these improved TS are equivalent to the existing 3

l TS definitions of the " component / system integrity".

Therefore, these changes are editorial in nature, that involve the movement or reformatting of requirements which remain applicable without affecting the technical content.

l 3.

The existing TS establish the secondary containment boundary while i

handling irradiated fuel inside the secondary containment. This has been i

clarified in the improved TS to include handling of fuel in the primary

)

and secondary containment. Since the secondary containment boundary completely surrounds the primary containment, handling irradiated fuel inside of primary containment also constitutes handling fuel inside the i

secondary containment. Thus, this change is administrative and is i

]

acceptable.

1

- ~. -.

1 i

i 89 i

4.

The improved TS do not include providing any " cross references". An example from the existing TS would be the reference to "Special Test j

Exceptions". These types of references serve no functional purposes and therefore, removal is purely an administrative preference in presentation which has no safety significance.

i 5.

The revised presentation of actions _in the improved TS do not propose to explicitly detail the most obvious option "to restore.... to operable status." This action, stated in the existing TS, is not repeated in the improved TS because this option is implicit in all conditions.

j Therefore, this provision is unnecessary and omitting this action is l

purely editorial.

1 6.

The improved TS LCOs, where appropriate, contain action note terminology such as... " Separate Condition entry is allowed for each (component, i

subsystem or system]." This guidance is implicitly utilized in the interpretation of the existing TS.

It is the presentation preference of the guidance in NUREG-1434 to explicitly state this instruction. The justification provided with improved TS in 1.3, Completion Times, further i

describes the usage of this terminology. This language is purely editorial in nature.

4 7.

The licensee proposes moving existing TS 3.6.1.6 leak-tightness requirements established pursuant to 10 CFR Part 50 Appendix J from the LCO to improved TS SR 3.6.1.1.1.

Existing TS 3.6.1.6 also describes the i

visual inspection done before each Type A containment leakage rate test, j

which is also described in 10 CFR Part 50 Appendix J.

These regulations 1

require licensee compliance, can not be revised by the licensee, and are l

addressed by direct reference in the TS. The TS need not duplicate the i

regulations. Therefore, the staff accepts the proposal to retain the TS requirement to meet the requirements of 10 CFR Part 50 Appendix J, as modified by approved exemptions, and eliminate from the TS the details i

which are also found in Appendix J.

The TS include the specified l

provisions required by Appendix J for inclusion in the TS.

j 8.

The valve alignment requirements in existing TS 4.6.3.3.a establish l

operability of the suppression pool cooling function of the RHR system, which the operator can actuate manually by repositioning valves and i

starting the RHR pump. The licensee proposes clarifying improved TS SR

{

3.6.2.3.1 to reflect the existing TS, which allow the suppression pool cooling system to be considered operable if the system valves can be i

aligned to the correct position.

4 j

9.

All drywell penetration valves and blind flanges required for drywell integrity by existing TS 3.6.2.1 are located either in the drywell or containment. Therefore, the 31-day frequency for surveillance of drywell penetrations for which existing TS SR 4.6.2.1.a is required, also meets the entry conditions of Note ** (a 92-day frequency), which applies to valves, blind flanges, or de-activated valves inside the drywell or containment. Existing TS SR 4.6.2.1.a is required for drywell penetrations not capable of being closed by operable automatic isolation y

l valves and required to be closed during accident conditions. Therefore,

4,4-.*u-4-Wah,-

m

---M+CJ2JJ 4

- - - --4 d4 _;

5 -

_.-dn

-au-a as-M

.-mi-**w 4

+

4+#p-A

., e

]

90 i

i the frequency for these penetratiens is not 31 days, but is actually 92 l

days as. proposed.-

10.

Improved LCOs 3.6.3.1, 3.6.3.2 -

6.3.3 all have a Required Action B.1 j

to verify by administrative means that the alternate hydrogen control.

i function is maintained. The Completion Time requires that this check be I

made in one hour and then repeated once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter. The repeat check every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is deleted in the improved TS because the LCOs controlling hydrogen mitigation systems have required conditions to l

j verify the status of redundant system (s) operability each time one of these LCOs is not met. This cross check accomplishes the same action as-i 4

repeated checking every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

i 11.

In a letter of October 22, 1993 (GNRO-93/00122)- the licensee proposed changes to the frequency for performing testing that verifies bypass j

leakage limits are met and changes to drywell airlock surveillance-l 1

requirements that would be controlled by the TS. The staff accepted the i

proposed changes in Amendment 119, February 16, 1995.

3

]

12.

In a letter of July 14, 1993 (GNRO-93/00002), the licensee proposed i

i changes to the hydrogen ignition TS. The staff accepted the proposed changes in Amendment 116, February 16, 1995.

1 j

The above changes are considered purely administrative changes in the statement of requirements in the improved TS and are, therefore, acceptable.

j l

B.

Relocated Requirements i

i In accordance with the guidance in NUREG-1434, the licensee proposed to

)

j relocate all or portions of the following existing TS within the improved TS:

Existina TS li.t].g 3/4.6.1.5 Feedwater Leakage Control System l

3/4.6.2.1 Drywell Integrity 3/4.4.2 Safety / Relief Valves 3/4.4.7 Main Steam Isolation Valves i

3/4.6.4 Primary Containment Isolation Valves i

3/4.6.6.3 Standby Gas Treatment System l

The more significant changes resulting from relocated items are as follows:

I 1.

The motor operated valves that are required in several existing TS 3/4.6 Specifications to be cycled in accordance with ASME Section XI for Inservice Testing (IST) are now specified as included into an inservice i

testing program located in the Administrative Control Section of the l

improved TS. Any testing details or exceptions as specified in the j

existing TS are relocated to the IST Program.

i 2.

The hardware requirements of existing TS 3/4.4.2.2, Safety / Relief Valves Low-Low Set Function, are proposed to be relocated to containment section i

l 4

=

91 improved TS 3.6.1.6.

The instrumentation and logic necessary for the Low-Low Set Function is addressed in improved TS 3.3.6.5.

3.

Existing TS 4.4.2.2 surveillance requirements for performing a Channel Functional Test, Channel Calibration and Logic. System Functional Test on the low-low set function pressure actuation instrumentation are proposed to be relocated to improved TS 3.3.6.5.

4.

Existing TS 3/4.4.7, " Main Steam Line Isolation Valves", is relocated and merged within the improved TS 3.6.1.3, " Primary Containment Isolation Val ves. "

)

i 5.

Existing TS 4.6.6.3'.b, c, d.2, d.5, e, and f surveillance requirements for the Standby Gas Treatment' System have been relocated to improved TS 5.5.7, Ventilation Filter Test Program. This is consistent with the format of NUREG-1434.

6.

A new LCO is added to the current licensing basis for the purpose of-implementing the Containment Systems portion of the improved TS.

Existing TS 3.6.2.1 included the former definition of Drywell Integrity, requirements for the drywell air lock, suppression pool, and isolation capabilities for drywell penetrations. The requirements of existing-TS 3.6.2.1 have been moved to improved TS 3.6.2.1 (suppression pool. average temperature), TS 3.6.2.2 (suppression pool water level), TS 3.6.5.1 (drywell operability), and TS 3.6.5.2 (drywell air lock).

Improved TS 3.6.5.3, Drywell Isolation Valves, was created to be consistent with the new definition of Drywell operability. The portion of Existing-TS 3.6.2.1 pertaining to drywell isolation valves is now in improved TS 3.6.5.3.

7.

The requirements for MSIVs in existing TS 4.4.7 state that LCO 4.0.4 does not apply to Modes 1 or 2 if the surveillance is completed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after meeting specified conditions. The improved TS 3.6.1.6 does not apply this restriction explicitly. The Bases for improved TS SR 3.0.4 state that'the proposed surveillance frequencies were written so that numerous exceptions to SR 3.0.4 are not required in the improved TS.

In this case, the details for the frequency are included in the ASME Section XI in-service testing program which is incorporated in 10 CFR 50.55a; therefore, the current TS exception is not required.

8.

The existing TS SR 4.6.6.3.d.3 that requires surveillance of the instrumentation portion of the standby gas-treatment system filter train start has been relocated to the improved TS 3.3.6.2.

9.

The hardware requirements for the low-low set function of the safety relief valves in existing TS 3.4.2.2 are stated in a separate LCO in TS 3.6.1.6.

The " hardware" addressed by this proposed LCO consists of the solenoid-actuated relief valves that support both the low-low set and relief functions of the safety relief valves. The existing TS also address related instrument operability requirements.necessary for the low-low set and relief functions. These requirements are relocated to TS 3.3.6.5 consistent with the format of BWR/6 Standard Technical

92 Specifications, NUREG-1434. These proposed changes are administrative because they are only changes to the format of the presentation.

The above changes are purely administrative changes in the location of the requirements in the improved TS, and are therefore acceptable.

4 In accordance with the guidance in the Final Policy Statement, the licensee has proposed to relocate or reorganize all or portions of the existing TS to other licensee-controlled documents.

j Existina TS Title 3/4.6.1.1 Primary Containment Integrity 3/4.6.1.2 Containment Leakage 3/4.6.1.3 Containment Airlocks 3/4.6.1.4 MSIV Leakage Control System i

3/4.6.1.5 Feedwater Leakage Control 3/4.6.1.8 Containment Average Air Temperature 3/4.6.2.2 Drywell Bypass Leakage 3/4.6.2.3 Drywell Airlocks 3/4.6.2.4 Drywell Structural Integrity 3/4.6.2.6 Drywell Average Air Temperature 3/4.6.3.1 Suppression Pool 3/4.6.3.2 Containment Spray 3/4.6.3.3 Suppression Pool Cooling 3/4.6.3.4 Suppression Pool Makeup System 3/4.6.4 Containment and Drywell Isolation Valves 3/4.6.5 Drywell Vacuum Relief 3/4.6.6.1 Secondary Containment Integrity 3/4.6.6.2 Secondary Containment Automatic Isolation 4

Dampers / Valves 3/4.6.7.1 Containment Hydrogen Recombiner 3/4.6.7.2 Containment and Drywell Hydrogen Ignition System 3/4.6.7.3 Combustible Gas Control Purge System l

The more significant changes resulting from relocated items are as follows:

1.

Appendix J of 10 CFR Part 50 delineates certain requirements that must be within the TS, and others that may be detailed within the Bases. The value of "P " is one requirement that is allowed to be presented in the Basesandtfiistestpressureisthereforerelocated. Any subsequent change to this test pressure would be controlled by 10 CFR 50.59 requirements and TS 5.5.11.

There are also numerous locations within i

existing TS 3/4.6.1.1 which detail the value "P,".

2.

The existing TS requirements for the operability of various instruments which provide indication of equipment condition or is used for equipment testing, excluding the post-accident monitoring instrumentation which is addressed specifically in Chapter 3.3 of this evaluation, will be relocated to the UFSAR or improved TS Bases.

Control of the availability of this instrumentation, and necessary compensatory actions if not available, is not relied upon for the associated system functions; this

=-

c i

i 93 instrumentation is not of the type required to be controlled by TS pursuant to the criteria in the Final Policy Statement.

3.

Existing TS 3/4.6 specifies in many of the specifications the details related to the system description, functional capabilities, and design of the system.(such as design of independent systems, number of components and flow path arrangements, etc.). This existing TS discussion is relocated to the improved TS Bases or UFSAR. The design limits and performance levels of components, subsystems, and systems will be relocated to the TS Bases or UFSAR. These details do not provide any useful or necessary information to the operator.

4.

Existing TS 3/4.6 specifies in many of the specifications the details of i

the methods for performing certain surveillance requirements. The details of the methods and-acceptance values for these continuity and system functional tests are relocated to the UFSAR and TS Bases. The values are system design values which were previously reviewed and approved by the. staff, and are also controlled by the design change procedures and 10 CFR 50.59.

5.

The existing TS_4.6.2.2 details relating to methods of performing surveillance test requirements and re-test frequency of drywell bypass i

leakage after two successive failures have been relocated to the improved i

TS Bases and UFSAR. The limits are retained in improved TS LC0 3.6.5.1.

The design features and system operation which dictate the methods are described in the UFSAR.

6.

The licensee proposes moving the detailed descriptions of the operability of the containment air lock (existing TS 3.6.1.3) knd drywell airlock operability (3.6.2.3) to the surveillances and bases to improved TS 3.6.1.2 and 3.6.5.2, respectively. Air lock interlock operability requirements are explicitly required for surveillances to verify air lock operability. The operability of this interlock ensures "when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed." The bases

. include discussion of the importance for both doors to normally remain closed is included in the bases.

If only one door remains closed, the safety design of the containment and its air locks ensures they would remain a sufficiently leak tight barrier for postulated events.

7.

The Main Steam Isolation Valve - Leakage Control System is-a manually actuated system. The instrumentation being tested by existing surveillance 4.6.1.4.d provides operational interlocks and automatic system shutdown for the system at various conditions. Many systems specified in the existing TS have system-internal operational functions that are not detailed within the Specification itself. The operability of these controls is incorporated by the operability of the system, and testing of these controls is considered part of the system functional test. Proposed SR 3.6.1.8.3 will inherently perform a functional test of the instruments and thus confirm the proper calibration of the channels.

The requirements will be relocated to the UFSAR and Bases l

l i

4 a

i 94 i

i The above relocated requirements relating to Containment Systems are not j

required to be in the TS under 10 CFR 50.36, and are not required to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety.. Further, they do not fall within any of the four criteria set forth in the Commission's Final Policy Statement, j

discussed in the Introduction above.

In addition,'the staff finds that i

sufficient regulatory controls exist under 10 CFR 50.59 or TS 5.5.11 to assure continued protection of the public health and safety. Accordingly, the staff has concluded that these requirements may be relocated from the TS to the j

licensee's TS Bases, UFSAR or plant procedures, as applicable.

C.

More Restrictive Requirements l

l By electing to implement the NUREG-1434 Section 3.0 Specifications, the licensee has adopted a number of more restrictive conditions than are required j

by the existing TS. The more significant conditions are the following:

j 1.

Existing TS 3.6.1.2, Action d. restricts heating up of the reactor coolant above 200Y with leakage above the applicable limit. This-4 existing TS action allows a start up and a control rod withdrawal from l

cold conditions to less than or equal to 200*F before requiring that the excessive leakage be corrected. The existing TS also does not contain specific action should the leakage be discovered during start up above a

i 200*F and before entering Mode 3.

The improved TS presentation and i

associated actions result in establishing and maintaining the reactor in a cold shutdown, all-rods-in condition until the leakage is corrected.

l The staff finds this improved TS requirement results in increased safety above the allowances in the existing TS action.

2.

Existing TS 3.6.1.6 is modified such that the Containment Structural Integrity requirement has been moved to the Primary Containment Operability LCO. The allowed time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to restore compliance with the existing TS on structural integrity, has been reduced to one hour for restoring Primary. Containment operability. The licensee accepts this i,

more conservative uniform Completion Time to avoid confusion in applying

]

different restoration times for the various components which comprise the primary containment boundary.

4 3.

Changes to existing TS 3.6.1.9 include proposed ras D.2 and D.3 that establish the requirement to periodically verify the isolated status and leak-tight status of the devices to isolate the containment purge system i

penetrations. These actions help ensure the safety of the plant.

l.

4.

Drywell Bypass Leakage, existing TS 3.6.2.2, restricts heating the reactor coolant above 200'F as a result of drywell bypass leakage above the applicable limit. This existing TS action allows start up and control rod withdrawal from cold conditions to less than or equal to i

200*F before requiring that the excessive leakage be corrected. The existing TS also does not contain specific action should the bypass leakage be discovered during start up above 200*F and before entering i

Mode 3.

The improved TS presentation and associated actions result in l

establishing and maintaining the reactor in a cold shutdown, all-rods-in,

d 95 t

~ condition until the bypass leakage is corrected. The staff finds this j

improved TS requirement results in increased safety above the allowances.

in the existing TS action.

4 d

5.

Drywell Structural Integrity, existing TS 3.6.2.4, requires that the drywell structural integrity.be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The licensee has proposed to change the existing LCO for "Drywell Structural-1 Integrity" to the existing LCO for "Drywell Integrity." The allowed time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to restore compliance with the existing TS on structural integrity has been reduced to one hour for restoring drywell operability.

The licensee accepts this more conservative uniform Completion Time to f

avoid confusion in applying different restoration times for the various j

components which comprise the drywell boundary.

6.

Existing TS 3.6.2.7 includes action for conditions when the penetration j

is not isolated, such as when both valves in the penetration are open.

i However, the TS do not require action if only one of two valves is not i

closed. Adding Condition A to the improved TS 3.6.5.3 would create such a requirement. This guidance is consistent with existing TS actions for isolating any opened penetration with inoperable isolation valves, and.

therefore does not introduce a safety concern.

7.

Suppression Pool, existing TS 3.6.3.1.b.3, limits the temperature rise of the suppression pool to 120'F when the MSIVs are closed following a j

scram. The improved TS limits the temperature to 120'F regardless of whether the MSIVs are open or closed since significant heat could still j

be added to the suppression pool. The required action is the

)

depressurization of the reactor. Additionally, this more restrictive action is appropriate even if the MSIVs are open, since there may be no heat rejected from the containment, as in the case of the loss of condenser vacuum. These required actions do not result in any operation which is not analyzed.

j 8.

Improved TS SR 3.6.4.2.1 requires the licensee to periodically verify the i

closure of each secondary containment isolation manual valve and blind i

flange that is required to be closed. These passive isolation devices have not previously been verified closed except as indicated by the 1

ability of the standby gas treatment system to create and maintain a i

vacuum. Therefore, this periodic verification is a more restrictive j

change.

i i

9.

Existing TS 3.6.6.2 states that each secondary containment. ventilation system automatic isolation damper shall be Operable. The proposed specification will delete the word " automatic" and therefore apply to all types of secondary containment isolation devices.

Since this is an added j

scope, the change is considered more restrictive.

);

The staff has reviewed these more restrictive requirements and concludes they result in an enhancement to the existing TS. Therefore, these more

{

restrictive requirements are acceptable.

o k

e

l*

1 i

i 96 D.

Less Restrictive Requirements j

The licensee, in electing to implement the NUREG-1434 Section 3.6 i

specifications, proposed a number of less restrictive conditions than are allowed by the existing TS. The more significant conditions are the following:

1.

Notes are added to several improved TS actions (e.g., TS 3.6.1.3, TS t

3.6.5.3) and to surveillance requirements (e.g., SR 3.6.1.3.3, SR 3.6.1.3.4, and SR 3.6.5.3.2) which permit the use of administrative controls to verify that various components and valves in high radiation 4

areas remain isolated, in the correct position, or locked closed.

l Administrative controls have been used in the improved TS for i

verification of valves / dampers in the TS for Primary Containment i

Isolation Valves, Secondary Containment Isolation Dampers / Valves, Drywell l

Isolation Valves, and for the door (s) in the Primary Containment Air.

l Lock (s) or Drywell Air Lock. The valves / dampers are initially verified as being in the correct position and/or that the isolation barrier is in place. The doors are initially verified lock closed.

Subsequently, i

access to the valves / dampers / door (s) is restricted during operation due i

to the controls placed on the areas with high levels of radiation. The staff concludes that the probability of misalignment of the isolation j

barrier is acceptably small, and that the change does not present a i

significant safety concern.

1 l

2.

Existing TS 4.6.1.1.b and Footnote ** to TS 4.6.2.1.a require the licensee to verify every 92 days that deactivated automatic valves in inaccessible or well controlled areas of the plant are locked, sealed, or otherwise secured. Accessible valves must be verified at a 31-day frequency. The licensee i

seal or otherwise secure" proposes to delete the requirement to " lock, from improved TS because it is more I

appropriately defined in the plant procedures. The bases fully describe i

the licensee's general methods to isolate a penetration, and plant j

procedures define the specific method for each penetration.

5 3.

The phrase in existing TS 4.4.2.2.b "by verifying actual or," has been added to several surveillances (e.g., SR 3.6.1.6.2) for the requirement i

to verify by simulated initiation that each system valve actuates to its correct position upon an automatic initiation signal. This allows using the occurrence of automatic initiation events from plant operations in l

fulfilling the surveillance requirements for various system components.

The operability of a component is demonstrated in either case; e.g., the valve can not discriminate between an " actual" or " simulated" initiation signal and its operability is demonstrated in either case.

4.

A note has been added to existing TS LCO 3.6.1.8 to allow inoperable 4

purge valves to be reopened under administrative controls. This change is consistent with the existing TS allowances for PCIVs (including these valves) under existing TS 3.6.4.

However, since these valves are also addressed under LCO 3.6.1.8 which does not contain this note, reopening these valves is not currently allowed if they are inoperable under LCO 3.6.1.8.

This change is consistent with NUREG-1434 and is acceptable

}

1

l

+

1 k-97 i

based on the limitation that the valves are to be under administrative-control while open. This ensures that the valves will be closed promptly -

)

in the event containment isolation is required.

i 5.

Containment Leakage, existing TS 4.6.1.2.f and h, requires the main steam isolation valves and hydrostatically tested valves to be tested on an 18 L

month frequency with a maximum extension of 25% or 4.5 months. This total interval of 22.5 months 4 provkied to allow for the scheduling of tests for these valves during each refueling outage. Appendix J of 10 CFR Part 50 allows Type B and C isolation valves to be tested on a i

interval of no greater than 24 months with no extensions. The Appendix J 3

frequency is also intended to provide for scheduling the valves to be i

tested at each refueling outage. The improved TS uses the Appendix J j

frequency limit of 24 months rather than the 18 month interval of the existing TS so MSIVs, hydrostatic tested valves, and the Type B and C 4

valves are all tested on the same schedule. This improvement prevents situations developing for unnecessary shutdowns; it extends the current interval by only 1.5 months; and it is consistent with the applicable i

regulations. No significant increase in leakage is likely to oc ur during this extended interval.

In addition, several studies document the minimal risk to the public posed by " unfiltered" containment leakage for l

j longer times and at rates far exceeding the current leakage limits.

i 1

6.

Existing TS 3.6.1.3 has been modified with a new allowance. This new Condition B in the improved TS pertains solely to an inoperable air lock l

i interlock mechanism.

If access into containment is desired, Note 2 i

permits an individual to be stationed at the air lock and dedicated to I

assuring that two doors are not open simultaneously and one door is re-i locked prior to leaving. This individual thus provides substantially the same level of protection as if the interlock mechanism were operable.

The condition further provides for periodic verifications that the air i

lock door remains locked until the interlock mechanism is returned to operable status and the condition is exited. The staff finds this is an i

equivalent required action which assures the air lock door is closed to

]

match the assumptions of the accident analyses.

7.

Existing TS 3.6.1;3 has been modified with a new condition. The improved TS has incorporated an actions Note #1 to permit entry through a closed i

or locked primary containment air lock door for the sole purpose of i

making repairs. An air lock with an inoperable outer door is fully accessible and the operable inner door maintains the containment operable. This Note would only apply to repairs made to an inoperable i

inner door of the air lock. Without this allowance, the inoperable door i

could prevent the overall air lock test from being performed and thus result in a plant shutdown from the inability to demonstrate the air lock operable. Additionally, it is the staff's preference to keep both doors operable in each air lock as an improvement on safety over just one 3

operable door locked closed.

It is possible to gain access to the i

inoperable inner door by entering the containment from the other air

- lock.

If this is not practical due to the length of travel distance or i

exposure considerations, then this Note would be utilized. This would result in the momentary loss of the primary containment boundary as the I

d a

't m

e e vevm -- - - - - - -

..m

, - ~ -

-en~

v r, v

1 l

98 l

outer door is opened for entry. The staff finds-this to be acceptable due to the low probability.of an event occurring that could pressurize the primary containment during the short time in which the containment 3

boundary is compromised.

8.

-Existing TS 3.6.1.3 Actions have been modified with a new condition in-the improved TS. Entry and exit through the primary containment air lock during normal operation is necessary to perform required surveillance, maintenance and inspections as well as allowing routine access for operational considerations such as chemistry sampling, Reactor Water j

Cleanup system operations, refueling preparations, etc.

If both air j

locks become inoperable and access is not allowed, a plant shutdown'would i

be forced in a short period of time due to failure to attend to these required activities. This Condition A, Note #2 is added to allow entry-for reasons other than repairs under strict administrative controls, which are detailed in the Bases, for a period of time not to exceed seven i

days.

In this one-time seven day period, an air lock must be returned to j

operable status or the forced shutdown must occur. The temporary loss of the containment boundary for brief times during this seven day period is i

4 judged by the staff to be acceptable. The risk associated with an event

]

occurring during the brief period of. time (not to exceed seven days)_is considered to be significantly less than the risk associated with a plant shutdown concurrent with plant' equipment that may not be'in a satisfactory operational condition.

9.

Existing TS 4.6.1.3.c requires verification, once every six months,'that only one door in each air lock can be opened at 'a time. A note to improved TS SR 3.6.1.2.2 proposes that this surveillance not be required to be performed unless the air lock doors are to be opened for a j

containment entry. Without this exception, the air lock doors would be j

required to be opened solely to perform this-interlock test. This

-j scenario would then also require the door seal test be performed within the next 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> creating unnecessary containment entries and requiring manpower for testing.

In the event the plant is ' utilizing one air lock for entries and maintaining one air lock idle, this surveillance would 1

impose an excessive testing requirement.

10.

Existing TS 3.6.1.4 requires both MSIV leakage control systems (LCSs) to be operable. The Action statement permits one LCS to be out of service i

for up to 30 days. However, if both MSIV-LCSs become inoperable, the

}-

facility is required to enter LCO 3.0.3.

In the improved TS, a period of 2

seven days is permitted to restore at least one subsystem to operable.

status; if operability is not established, an orderly plant shutdown is i

i required. The longer Condition B Completion Time permits additional repair time to restore a subsystem to operable status. These relaxations are based upon these potential leakage paths not being a direct path to the environment such as via leaking containment purge valves. This means the MSIVs are the first line of defense to isolate and minimize containment leakage. The second line of defense is the MSIV-LCS. The i

most important function of the MSIV-LCS is in the case of a seismic event which disables or fails the MSIVs. The LCS would function in lieu of the MSIV. The MSIV LCS is judged to be of comparably low safety significance i

_.y i

i 99 i

}

because the subsystems are required to filter only a small portion of the I

l complete potential containment leakage.after a design basis accident.

j Several studies have documented the minimal impact of increased unfiltered containment leakage. Among these are NUREG-1273, Technical 1

Findings and Regulatory Analysis for Generic Safety-Issue II.E.4.3, i

" Containment-Integrity Check," and NUREG/CR-3539, " Impact of Containment 1

Building Leakage on Light-Water Reactor Accident Risk." These documents indicate that leakage rate increases that are significantly in excess of s

the allowed MSIV leakage rates would not result in significant increase i

in risk to the public. Therefore, the staff has specified an out-of-i service time that is commensurate with the actual need for the MSIV-LCS.

In the judgement of the staff, these changes are acceptable.

i i

11. Existing TS 3.6.1.5 requires both feedwater (FW) leakage control systems l

(LCSs) to be operable. The TS permit one LCS to be inoperable for up to.

30 days. However, if both FW LCSs become inoperable, the facility is d

l required to enter LCO 3.0.3.

Condition E in improved TS 3.6.1.8 permits 4

a period of seven days to restore at least one subsystem to operable q

status; if operability is not established, an orderly plant shutdown is required. The longer Condition B completion time permits additional j

repair time to restore a subsystem to operable status. The action s

completion times assume that these potential. leakage paths such as leaking containment purge valves are not direct paths-to the environment.

l The specified out-of-service time is commensurate with the actual need i

for the FWIV LCS; therefore, the change is acceptable.

i I

}

12. The licensee proposes changing existing TS 3.6.1.8 to adjust the containment average air temperature limit to that limit assumed in the 4

accident analysis. The present temperature limit (90 degrees 'F) was established to ensure continued capability of equipment in the i

i containment. The new limit of 95 degrees in improved TS 3.6.1.5 is the i

j actual temperature limit assumed in the accident analyses. The 90-degree limit for equipment qualification will still be monitored by plant l'

procedures, but will not be the required limit.

i 13.

Existing TS 4.6.1.9.2 requires a testing interval for the 36-inch vent and purge valves of 92 days. The frequency for performing this j

surveillanc'e has been extended to 184 days in the improved TS.

If the valve is opened any time within the frequency interval then this surveillance must be performed once within 92 days after the opening of the valve. This extension is consistent with the NRC guidance provided in accordance with the resolution of Generic Issue B-20, " Containment j

Leakage Due to Seal Deterioration."

s I

14.

Existing TS 3.6.2.1 requires that a drywell isolation valve be restored to operable status in one hour regardless of the configuration of the drywell penetration flow path. TS 3.6.2.7 requires the drywell vent and purge valves to be closed immediately. The improved TS.3.6.5.3, "Drywell i

Isolation Valves," defines new Conditions A and B based upon the number i-of isolation valves in the penetration flow paths.

This definition i

permits the alternate action of isolating the flow path rather than restoring the operability of the affected valves. The completion time l

j t

4

8 O

l l

i.

100 i

i for this allowance. is extended to four hours because the restoration of -

the drywell boundary would ensure.the ability of the drywell boundary to withstand the pressure loads of a design basis accident and also permit' 4

the continued operation of the plant.

15. Existing TS SR 4.6.2.2.c, SR 4.6.2.3.a SR 4.6.2.3.b.1 and Footnote #

state that the provisions of.LC0 4.0.2 do not apply to drywell bypass leakage and drywell airlock tests, respectively. -The improved TS removes-this restriction, which prevents use of the surveillance frequency extension of 25 percent in improved TS _SR 3.0.2.. Applying this restriction is unnecessary since the general. provisions of TS establish 4

I frequency extensions for scheduling surveillance activities to a period when plant conditions are suitable-for the' surveillance.

16. The existing TS 4.6.2.2 requirement to have the schedule for drywell a

bypass leakage test reviewed and approved by the Commission is deleted 4

from the TS as no longer necessary. This requirement is not included in NUREG-1434. The Commission will be informed of any bypass test failure j

by the Licensee Event Reporting (LER) system required in 10 CFR 50.73.

j The. Commission can review any changes to the test schedule or require any l

changes to the test schedule at that time in accordance with regulations..

17. Existing TS 3.6.2.3 has been modified. The improved TS has added a Condition B to address an inoperable air lock interlock mechanism which restricts entry into drywell. The assumptions of the accident analyses are maintained and operation is allowed to continue when one operable air lock door in an air lock can be maintained closed and verified to be j'

locked to assure it remains closed.

In the event drywell access is j

' desired, it is allowed under strict administrative controls which provide i

a level of assurance substantially equivalent to an operable interlock i

' mechanism. These controls require a dedicated individual to assure that two doors are r.ot open simultaneously and one door is re-locked prior to j

leaving.

i j

18.

Existing TS 3.6.2.3 has been modified. The improved TS has incorporated j

an Actions Note #1 to permit entry through a closed or locked air lock i

door for the sole purpose of making repairs. An air lock with an inoperable outer door is fully accessible and the operable inner door i

maintains the Drywell operable. This Action Note situation would only apply to repairs made to an inoperable inner door of the air-lock.

i Without this allowance, the inoperable door could prevent the overall air i

lock test from being performed and thus result in a plant shutdown from the inability to demonstrate the air lock operable. Additionally, it is the staff's preference to keep both doors operable in each air lock as an j

improvement on safety over having just one operable door locked closed.

I 19.

Existing TS 3.6.2.3 has been modified with a new condition in the improved TS.

Entry and exit during normal operation may be desired to perform maintenance and inspections.

If an air lock' door becomes i

inoperable and access is not allowed, a plant shutdown would be forced in a short period of time due to an inability to attend to necessary activities. This Condition A Note #2 is added to allow entry for reasons

-..-w

~

a.

a i

i 101 i

other than repairs, under strict administrative controls which are i

detailed in the Bases, for a period of time not to exceed seven days.

In this one-time seven day period, an air lock must be returned to operable 4

i status or the forced shutdown must occur. The temporary loss.of the j

drywell boundary for brief times during this seven day period is judged by the staff to be acceptable. The risk associated with an event occurring during this brief period of time (not to exceed seven days) is considered to be significantly-less than the risk associated with a plant shutdown concurrent with plant equipment that may not be in a.

satisfactory operational condition.

j 20.

Existing TS 4.6.2.3.b requires a_ barrel test every six months. The.

l:

drywell air lock is typically tested similar to the primary containment air locks; however, the drywell air lock is not a direct leakage path-i from the primary containment and, therefore, Appendix J requirements do not apply.

In addition, the drywell airlock does not experience frequent usage due to the radiation and temperature in the drywell. The improved TS interval requirement of a barrel test has been extended to 18 months.

1 The staff finds the verification of the seal leakage rate after each use of the drywell is sufficient to assure sealing capability and thus justifies the scheduling of this test at refueling / outage intervals.

2

21. The licensee proposed clarifying existing TS 4.6.2.3.c in improved TS SR i

3.6.5.2.3 to add a note to limit the timing of the surveillance for the j

drywell air lock door interlock mechanism so that it is required only before the drywell is entered. This exception eliminates many activities i

i associated with the need to open the door just to do the surveillance.

4 It also eliminates the need to do the seal leakage rate test before i

1 opening the door. The note will prevent unnecessary drywell entries, reduce cycling of the door seals, reduce staff constraints and minimize exposures during testing.

l l

l 22.

Existing TS 3.6.2.7 Action b., and 4.6.2.7.2 specifies a maximum time limit of five hours per year for the opening of either the 6 or 20 inch i

drywell vent and purge supply and exhaust valves. The improved TS l

proposes to remove the specified hour limitations, to be replaced with the specific criteria defined in the note to SR 3.6.5.3.1.

The previous time limits were not based on any legal requirements but were derived I

from engineering judgement and early plant operating experience.

[

Additional plant operating history has shown that these valves would have been open under the new criteria for cumulative periods significantly less than the previously allowed cumulative times, and the limits are 3

therefore unnecessary.

1 23.

Existing TS 3.6.3.1, Action a. allows one hour to restore the' suppression 4

j pool water level. This allowed outage time was based upon engineering-judgement. An unanticipated change in suppression pool water level requires addressing the cause of the problem and the selection of the

]

appropriate water system to align in order to restore the level. This i

must be accomplished without creating a new problem in the plant

}

personnel's haste to remedy the first problem.

The improved TS proposes to extend this restoration time to two hours so that action may occur:

f 1

l l

102 i

rapidly but with better assurance of the appropriateness of the actions to be completed.

24. Existing TS 3.6.3.1 Action b.2 requires that one RHR loop be placed in

]

the cooling mode when the temperature exceeds 110'F. Maximum allowable.

sugpression pool temperature during normal power operations is limited to

.95 F and severe operating restrictions are imposed when this temperature is exceeded. Therefore, it is highly likely that when.the~ pool 4

temperature exceeds 95'F, efforts would already be under way to monitor and reduce the temperature to the extent practical. However, plant operations could result in dedication of RHR subsystems to functions 1

j other than suppression pool cooling. Also, requiring that an ECCS subsystem be placed in the suppression pool cooling mode is. unnecessary j

because the operator is precluded from choosing a more prudent course of 1

i action. The staff finds the plant operating procedures contain the l

j appropriate actions to direct the activities necessary to return the j

plant to a safe, stable configuration.

)

25.

Existing TS 4.6.3.1.b.2 and 3 requires suppression pool temperature verification once per hour.when the temperature is between 95'F and 4

j 110*F. The specification also requires hourly verification that Thermal 1

Power is less than or equal to 1% Rated Thermal Power. The improved TS i

requires this same action (Required Action A.1) only if Rated Thermal Power is greater than or equal to 1%. Also, the existing TS 4.6.3.1.b.3 requires pool temperature verification every 30 minutes following a scram l

when pool temperature is between 95'F and 120*F whereas the improved TS requires this same action (Required Action D.2) only between 110*F and l

l 120*F. Due to the severe operating restrictions imposed when pool i

temperatures exceed 95'F, it is highly expected that plant operators would already be taking alternative actions to stabilize the suppression

}

pool temperature if it exceeds 95'F.

The staff finds there is minimal i

significance to modifying these existing actions since the operator knows l

the current power level at all times.

26. Existing TS 3.6.3.2, Action a. allows one containment spray loop to be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> before a plant shutdown is required. The l

improved TS extends this restoration time to seven days. The same components of the LPCI ECCS subsystem covered in the improved TS 3.5 1

includes a seven day allowed outage time. Therefore, the allowed outage time of seven days is permitted as being consistent with what is allowed i

for an inoperable ECCS subsystem in improved TS Section 3.5.

I

27. The existing TS 4.6.3.2 requirement is modified by adding a note to t

improved TS SR 3.6.1.7.1 to avoid conflicting with existing TS 3.4.9.1, j

which requires two RHR-shutdown cooling subsystems be operable with one in operation. This note permits the containment spray system function to i

be considered operable if aligned to the shutdown cooling function when below the cut-in permissive pressure. This exception is only permitted if the containment spray function can be performed on manual reali and the containment spray is not inoperable for some other reason.gnment I

i

o 103

28. Suppression Pool Cooling, existing TS 3.6.3.3, Action a allows one suppression pool cooling loop to be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> before a plant shutdown is required. The improved TS extends this restoration time to seven days. The same components of the LPCI ECCS subsystem are permitted a seven day allowed outage time in improved TS 3.5.

Therefore, the allowed outage time of seven days is permitted here as being consistent with what is allowed for an inoperable ECCS subsystem in improved TS Section 3.5.

29. Existing TS 3.6.3.4, Action a. allows one suppression pool makeup subsystem to be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> before the plant shutdown is required. The improved TS extends this restoration time to seven days, similar to what is required in the improved TS 3.5 ECCS specifications.

There is a low probability of a design basis accident occurring during this extended seven day period. Therefore, the allowed outage time of seven days is permitted here as being consistent with what is allowed for an inoperable ECCS subsystem.

30.

Existing TS 3.6.4 requires the licensee to maintain at least one isolation valve operable in each open penetration and restore the inoperable valve or isolate the penetration in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if one or more containment or drywell isolation valves is inoperable. The improved TS extends this completion time to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for drywell isolation valves and applies the time to those drywell isolation valves with a cross sectional area less than the equivalent assumed bypass leakage area.

In this case, j

the limiting event would still be within the bounds of the safety analysis. Allowing an extended restoration time to avoid a plant i

transient caused by the forced shutdown is reasonable for the probability of a drywell pressurization event and would not significantly decrease safety.

31.

For PCIVs, existing TS 3.6.4, Action b. requires that each containment penetration having an inoperable PCIV be isolated by use of at least one deactivated automatic valve secured in the isolated position. The improved TS defines an acceptable isolation mechanism to include a check valve with the flow through the valve secured.

This condition provides an acceptable boundary and an acceptable control to ensure that the boundary is not inadvertently opened. However, this allowance has limitations, for example, the allowance is not permitted if both PCIV's in a penetration are inoperable. Under these limitations, use of a check valve in lieu of a deactivated automatic isolation valve is acceptable.

32.

For MSIVs, existing TS 3.4.7, Action a. and for other valves existing TS 3.6.4 require maintaining one isolation valve operable in a penetration that is open; otherwise, an immediate shutdown is required. The improved TS provides Completion Times for restoring one valve to operable status or isolating the penetration that are consistent with the Completion Times for Conditions when the primary containment is inoperable.

Instead of "immediately", a one hour Completion Time is provided for a primary containment boundary valve and four hours is provided for a secondary containment boundary valve. The staff finds the proposed change will

~.

,[-

i i

i i '

104 e

i provide for a consistent response for these various containment l

degradations and will not adversely affect plant safety.

4 33.

For MSIVs, existing TS 3.4.7, Action a.1.b requires the affected main steam line be isolated by use of a deactivated MSIV in the closed position. The improved TS additionally permits isolation with a manual l

valve or blind flange in the combined LCO for all PCIVs._ For the specific case of the main steam lines, isolation could be provided by safety-related Class 2 manual valves which are located downstream of the i

MSIVs. The staff finds the ability to utilize the valves downstream of the outboard MSIV to be an acceptable isolation barrier since it meets the acceptance criteria of not being affected by'a single active failure.

[

34. - Footnote

• to existing TS 3.6.4, action b, permits isolation valves l

except MSIVs to be reopened periodically under administrative controls.

i Improved TS 3.6.1.3 would also allow valves used to isolate main steam i

line penetrations to be reopened. This is consistent with all other penetrations. Opening the valves used to isolate the main steam line penetrations may be necessary to prepare the line for a return to service

.(draining or pressurizing the line). The restriction can be removed because these isolation valves are a boundary similar to other j

containment isolation valves with this allowance.

35. For PCIVs, existing TS 4.6.4.2 requires that each automatic primary l

containment or drywell isolation valve be demonstrated operable during Cold Shutdown or Refueling at least once per 18 months by verifying that on an isolation test signal each automatic primary containment isolation valve actuates to its isolation position. The improved TS permit the i

scheduling of these tests when plant conditions are appropriate to i

perform the test in accordance with plant procedures. This action is consistent with the other TS that state an interval but do not dictate i

the plant conditions for the surveillance. The removal of this l

restriction from the TS has been determined by the staff as not i

i jeopardizing safe plant operations.

I 36.

Existing TS 3.6.5, action d, requires closing the individual open vacuum j

relief valves, whereas improved TS 3.6.5.6 requires the closing of an j

open penetration for each subsystem. Open individual valves in a i

penetration will continue to be.a path for vacuum relief; however, an i

i open penetration would defeat the original pressure suppression purpose i

of the drywell. By closing one of the subsystem valves, the licensee 4

would maintain drywell isolation. The valve should not be deactivated-i because the vacuum relief safety function is more important than the isolation function in this case.

37.

Existing TS 3.6.5, Action d. requires one hour to cloce at least one open vacuum relief valve in each penetration whereas the improved TS allows 1

four hours to complete this action.- The one hour time is still retained in the improved TS 3.6.5.1 if the drywell is inoperable. This change is made here to implement consistent Completion Times as is done similarly.

l for all. drywell isolation valves.

)

I

i 105

38. Condition D in improved TS 3.6.5.6 modifies existing TS 3.6.5 to allow three subsystems of drywell vacuum relief to be inoperable for.up to 72-i hours.

If all four subsystems are inoperable, an immediate plant shutdown is required. This new condition is consistent with the i

different combinations of subsystems allowed to be inoperable in action c to existing TS 3.6.5.

However, the change is a less restrictive requirement because this previously undefined condition would require the existing TS LCO 3.0.3 limitations.

k 39.

Existing TS 3.6.6.2 is modified in the improved TS, which contains i

Actions Note 1 and Note 2 to SR 3.6.4.2.1.

The first note permits the use of administrative controls to verify that secondary containment isolation valves, dampers, and blind flanges are in the closed position.

j This note would be applicable for those devices located in high radiation

?

areas. The second note permits the opening of manual isolation devices 2

required to be kept closed under action statements. The addition of j

these notes provides the same operational flexibility granted in the improved TS to the Primary Containment Isolation Valves and is acceptable.

q i

40. Existing TS 3.6.6.2 is modified in the improved TS, which contains Condition B for two inoperable valves / dampers in a penetration flow path.

The existing TS requires an immediate shutdown whereas the improved TS i

provides a four hour Completion Time to restore the valve / damper operable or isolate the penetration. This Completion Time is consistent with the existing TS time allowed when secondary containment is inoperable. This change also provides consistency with the changes to the actions for various primary and secondary containment boundary degradations.

I 41.

Existing TS 4.6.6.2.b requires that at least once per 18 months during l

Cold Shutdown or Refueling, verification must be made that each secondary j

containment automatic isolation damper actuates to its isolation position following.a test signal. The improved TS does not specify the plant i

conditions for performance of these tests thus permitting greater i

operational flexibility. This action is consistent with the other technical specifications that state an interval but do not dictate the plant conditions for the surveillance.

The staff finds that the rencval 1

of this restriction from the technical specifications does not jeopardize l

safe plant operations.

42.

Existing TS 3.6.6.3 is modified in the improved TS. A new Condition C is added to improved TS 3.6.4.3 to enable fuel handling operations to continue in the event one standby gas treatment system cannot be returned to operable status within seven days. This condition allows the i

remaining operable SGTS subsystem to be placed in operation while fuel-handling operations. continue. Since only one subsystem of SGTS is i

required during any accident, risk of the subsystem not being available i

is significantly reduced if it is already running.

Correspondingly, if i

this subsystem can not be started or stops, then all fuel handling operations must cease immediately.

p i

l l

i

. - -. - - _ =.

4 106 l:

1

43. Existing TS 3.6.7.1 is subject to the mode change restrictions'of j

Specification 3.0.4 when one of the two required containment hydrogen 1

recombiners is inoperable.

Improved TS 3.6.3.1 includes a statement to i

indicate that TS 3.0.4 is not applicable.

Existing TS 3.6.7.1 also only addresses a single inoperable hydrogen recombiner.

Improved TS 3.6.3.1

)

provides a new condition B if both containment hydrogen recombiners are inoperable. Condition B includes a period of seven days to restore one 1

of the two recombiners to operable status. This time is contingent upon i

verifying, by administrative means, that an alternate hydrogen control function is_ operable and available to back up the recombiners.

If a recombiner cannot be restored to operable status within seven days, a plant shutdown is required.

j Unlike many other mitigation systems (e.g., ECCS and RHR), any failures i

in the hydrogen recombiner system will not initiate plant transients or l

accidents.

In addition, the design basis for the hydrogen control system i-required under i 50.44 assumes severe core damage beyond the capability l

j of the ECCS and engineered safety features design bases. Because of the j

lower likelihood of events requiring this hydrogen control capability and i

the time available for operator action to cope with hydrogen accumulation, as compared to the risks associated with transients caused by forced plant shutdowns and related plant operations, the staff has concluded that there is no need for the inode change restriction and that a 7 day completion time is justified for a complete loss of function j

(i.e., failure of both recombiners).

a

44. Existing TS SR 4.6.7.1.a requires two functional tests of the hydrogen recombiners. One occurs every eighteen months at the design basis recombiner operating temperatures and a second is conducted at reduced temperatures every six months. This second test is proposed to be i

eliminated as is recommended in NUREG-1366, " Improvements to Technical i

4 Surveillance Requirements." This is based on the redundancy provided for the hydrogen control function, the system's historical high reliability, 2

and the delayed nature of the requirements for starting the system. The l

deletion of this redundant functional test does not have an adverse i

effect upon safety. Therefore, the staff concludes the changes are acceptable for the foregoing reasons and for the reasons described under a

i item 48, above.

45.

Existing TS 3.6.7.2 is subject to the mode change restrictions of l

Specification 3.0.4 when one of two required mixing system becomes inoperable. The improved TS has included a statement to indicate.that Specification 3.0.4 is not applicable when one containment /drywell i

hydrogen mixing system is inoperable.

Existing TS 3.6.7.2 also'only j

~

addresses a single inoperable containment /drywell hydrogen mixing subsystem. The improved TS provides a new Condition B if both subsystems i

are inoperable. Condition B includes a period of seven days to restore one of the two containment /drywell hydrogen mixing systems to operable

^j status. This time is contingent upon verifying, by administrative means, that an alternate hydrogen control function is operable and available to backup the containment /drywell hydrogen mixing system.

If a i

I

e l

107 containment /drywell hydrogen mixing system can not be restored to operable status within seven days, a plant shutdown is required.

The staff has concluded that there is no need for the mode change restriction and that a 7 day completion time is justified for a complete loss of function (i.e., failure of both recombiners) based on the reasons described above under item 48, above.

I

46. Existing TS 3.6.7.3 is subject to the mode change restrictions of Specification 3.0.4 when one of two required hydrogen subsystems become inoperable. The improved TS includes a statement to indicate that Specification 3.0.4 is not applicable when one hydrogen igniter subsystem is inoperable. Existing TS 3.6.7.3 also only addresses a single inoperable containment /drywell hydrogen igniter subsystem. The improved TS provides a new Condition B if both hydrogen igniter subsystems are inoperable. Condition B includes a period of seven days is allowed to restore one of the two hydrogen igniter divisions to operable status.

This time is contingent upon verifying, by administrative means, that an alternate hydrogen control function is operable and available to backup the hydrogen ignition systea.

If the hydrogen igniter division can not be restored operable within seven days, a plant shutdown is required.

The staff has concluded that there is no need for the mode change restriction and that a 7 day completion time is justified for a complete l

loss of function (i.e., failure of both recombiners) based on the reasons described above under item 48, above.

47. Any time the operability of an existing TS 3.6 system or component has been affected by repair, maintenance, or replacement of a component, as discussed in the Bases for SR 3.0.1, post maintenance testing is required to demonstrate operability of the system or component. Therefore, the licensee proposes deleting explicit post maintenance SRs from the specifications.

48. The licensee proposes deleting system operational details from the existing surveillance 4.6.5.b.3.b and c.

These controls must be operable for the system to be operable, and they are tested in the functional tests of the individual valves (improved TS SR 3.6.5.6.2) and the setpoint verifications of these valves (improved TS SR 3.6.5.6.3) which confirm the proper calibration of the respective channels.

49. The information required in existing SR 4.6.1.6.2, "Special Report," is ilso required by 10 CFR 50.73 and Appendix J to 10 CFR Part 50. The timing of the report as required by regulations changes with the seriousness of the containment degradation. Appendix J to 10 CFR Part 50, EV.B., requires only that this information be submitted with the ILRT Report.

If the principal safety barrier (the containment) is seriously degraded, a 30-day report is required by 10 CFR 50.73. The Special Report requirement in existing TS duplicates reporting requirements in the regulations and can be deleted.

108 The above less restrictive requirements have been reviewed by the staff and have been found to be acceptable, because they do not present a significant safety question in the operation of the plant. The TS requirements that remain are consistent with current licensing practices, operating experience and plant accident and transient analyses, and provide reasonable assurance that the public health and safety will be protected.

3.7 Plant Systems A.

Significant Administrative Changes.

In accordance with the guidance in the Final Policy Statement, the licensee has proposed administrative changes to the existing technical specifications

(

(TS) to bring them into conformance with the improved TS. These changes are as follows:

l 1

1.

Existing TS 3.7.2 Actions has an implied action for which improved TS l

3.7.3 Condition D directs entry into LCO 3.0.3 if both subsystems are i

inoperable in Mode 1, 2, or 3.

This avoids confusion as to the proper action if in Mode 1, 2, or 3 and simultaneously in a special condition such as handling irradiated fuel in the Fuel Building.

Since this action results in the same action as the existing TS, this change is i

administrative in nature and is acceptable.

2.

The existing TS 3.7.2 Action b.2 to immediately suspend OPDRVs is not l

possible in all plant conditions.

If not performed immediately, the i

result is non-compliance with the TS and a requirement for a LER.

Improved TS 3.7.3 Actions C.2.3 and E.3 more appropriately present the intent of the existing TS actions. These actions impose a requirement to initiate actions to suspend OPDRVs.

Implicit in this requirement is the understanding that best efforts to suspend OPDRVs must continue until

+

finished. However, if the suspension of OPDRVs cannot be accomplished immediately, no LER is required. As an enhanced presentation of the existing intent, the change is administrative and is acceptable.

3.

In implementing existing TS SR 4.7.2, a new surveillance requirement (improved TS SR 3.7.3.2) is added to clarify that the tests of the

. Ventilation Filter Testing Program must also be completed and passed for determining system operability. Since this addition maintains existing TS requirements, it is an administrative change and is acceptable.

4.

The licensee proposes modifying existing TS 3.7.2 by allowing the exiting of the applicable mode (s) as an alternative to the 7-day time to restore an inoperable CRFA subsystem to operable status. This action is implied in existing TS 3.0.1 and improved TS LCO 3.0.1.

Therefore, the allowance i

in improved TS 3.7.3 to exit the applicable modes is administrative and acceptable.

5.

The existing TS SR 4.7.2 a. for the control room emergency filtration subsystems heaters requires the heaters to be operable, while improved TS SR 3.7.3.1 requires the heaters to be operating to eliminate moisture

l-109 from the adsorbers and the HEPA filters. The licensee proposes this change to test the ability of the heaters to operate. This change will not affect the results of current testing practices and therefore is administrative.

j-6.

The existing TS 3.11.2.7 alternatives to compliance with the LCO are to j-exit the condition or to exit the applicable mode (s).

Improved TS 3.7.5 3

allows continued operation.' The improved TS 3.7.5 Condition B allows the isolation of the steam jet injectors, an action that curtails further release and has the same completion time as exiting _the mode (s).

Although this change is less restrictive because it allows continued operation, it maintains a level of safety equal to exiting the mode and j

is therefore administrative.

I 7.

Footnote ** to existing TS 3.7.1.1 Action a.2 requires the reactor i

coolant to be maintained at a temperature as low as possible if a cold shutdown temperature can not be attained.

Footnote ** contains no additional restrictions on the operation of the plant and could be i

interpreted as a relaxation of the existing TS 3.7.1 requirements to 1

achieve cold shutdown. This required action is also required by existing TS 3.4.9.1 and the actions of improved TS 3.4.9.

Therefore, removal of this footnote is an administrative change in the placement of TS j

requirements.

1

(

8.

The format of the improved TS does not include cross references because 1

the applicable improved TS adequately prescribe the remedial actions j

t without such additional direction. Therefore, the existing TS 3.7.1.2 LCO direction that the licensee "take the action required by..." or declare the diesel generator inoperable, which serves no safety purpose.

Therefore, removing this direction is purely an administrative difference l

in presentation.

i Since these requirements result in the same limits as the current requirements, the changes are administrative in nature and are acceptable.

l B.

Relocated Requirements l

In accordance with the guidance in the Final Policy Statement, the licensee 3

has relocated all or part of the following existing TS sections to other '

I i

locations in the improved TS:

i Existina TS Title 3/4.7.2 Control Room Ventilation System j

3/4.7.3 Reactor Core Isolation Cooling (RCIC) System 3/4.7.7 Liquid Storage Tanks i

3/4.7.8.1 Offgas-Explosive Gas Mixture j

3/4.7.8.2 Offgas-Noble Radinactivity Rate i

The more significant changes resulting from relocated items are as follows:

j j

1.

All or part of the technical contents of existing TS 4.7.2 b.,

c., d.1, i

d.2, d.3, e and f. which specify ventilation filter testing and

=

4 1-i a

110 J

surveillance requirements, are relocated to improved TS Section 5.5.7, Ventilation Filter Testing Program.

i 2.

The entire existing TS 3.7.3 technical contents are moved to improved' l

TS 3.5.3, RCIC System.

These~ requirements are included in the TS as required by 10 CFR 50.36a. The i

changes are considered administrative changes' involving only a change in the j

location.of the requirements in the TS and are, therefore, acceptab' e.

i In accordance with the guidance in the Final Policy Statement, the licensee has relocated all or part of the followir.g existing TS sections to'other licensee-controlled documents:

i Existina TS Ijilg j

j; 3/4.7.1.1 Service Water System j

j[

3/4.7.1.2 High Pressure Core Spray Service Water System 3/4.7.1.3 Ultimate Heat Sink i

3/4.7.2 Control Room Emergency Filtration System i

3/4.7.4 Snubbers i-3/4.7.5 Sealed Source Contamination l

3/4.7.6 Main Turbine Bypass System-l 3/4.7.8 Offgas - Explosive Gas Mixture 3/4.7.9 Spent' Fuel Storage Pool Temperature i

3/4.7.10 Flood Protection 3/4.9.9 Water Level - Spent Fuel Storage and Upper Containment Fuel Pools l

3/4.11.2.7 Main Condenser j

The more significant changes resulting from relocated items are as follows:-

4 t

i

'1.

The details in existing TS relating to system design and purpose are j

relocated to the improved TS Bases, in accordance with the format of the 3

improved TS. The UFSAR also describes the design features and system i

)

operation. The applicable provisions in improved TS Chapter 5 will i

control changes to the Bases.

1 2.

Existing TS 3.7.1.1, Standby Service Water System, and TS 3.7.1.2, Ultimate Heat Sink, include Applicability in all operating modes including when irradiated fuel is being handled in the primary or i

secondary containment. The requirements for Operability of these systems 1

in non-operating Modes, i.e., Modes 4, 5 and any special conditions such I

as when iandling irradiated fuel in primary or secondary containment, i

have been relocated to UFSAR. Since these systems are support systems for other required equipment with their own specifications, the definition of operability will provide sufficient requirements to assure these systems can adequately perform their required support function.

j The design features and system operation for the supported systems are 4

also described in the UFSAR.

i' 3.

Details of the methods for performing SR 4.7.1.1 b.; SR 4.7.1.2 b.; parts-i of SR 4.11.2.7.2.a and 4.11.2.7.2.b are relocated to the in accordance i

I L

. o 111 with the format of the improved TS. The design features describing detailed surveillance test methods and plant conditions are included in the UFSAR and TS Bases. The staff concludes that this descriptive information is not necessary for inclusion in the improved TS.

Additionally, the provisions. in improved TS Chapter 5 will control i

changes to the Bases and 10 CFR 50.59 will control changes to the UFSAR.

4.

Existing TS 3.7.4, " Snubbers," states that all snubbers shall be i

operable. Snubbers are passive devices used for supporting piping systems and the associated TS action statement only requires that an inoperable snubber be replaced or repaired within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The l

surveillance requirements for snubbers is that they be periodically i

examined under the inservice inspection program. The requirements of j

4 4

existing TS 3.7.4 that all snubbers be operable are requirements that do not impact reactor operation, do not identify a parameter that is an initial condition assumption for a Design Basis Accident (DBA) or transient, do not identify a significant abnormal degradation of the reactor coolant pressure boundary, and do not form part of the primary success path which functions or actuates to mitigate a design basis accident or transient. Therefore, the requirements specified in the existing TS have been relocated to the UFSAR or TS Bases and will be controlled in accordance with 10 CFR 50.59.

1 5.

Existing TS 3.7.5, " Sealed Source Contamination," states that sealed sources containing radioactive material shall be free of a specified 4

removable contamination. The associated action Ltatement requires that if the removable contamination exceeds limitations, the sealed source shall be either decontaminated or disposed of. The limitations expressed in this TS do not impact reactor operation, do not identify a parameter 1

which is.an initial condition assumption for a DBA or transient, do not identify a significant abnormal degradation of the reactor coolant pressure boundary and do not provide any mitigation of a design basis i

event. Therefore, the requirements specified in the existing TS have been relocated to the UFSAR or TS Bases and will be controlled in l

accordance with 10 CFR 50.59.

4 6.

Existing TS 3.9.9 requires that at least 23 feet of water shall be maintained over the top of irradiated fuel assemblies stored in the spent fuel storage and upper containment fuel pool racks.

The associated action statement restricts the movement of fuel assemblies and other crane operations with loads when the above water level is not met. The restrictions concerning " crane operations with loads" has been relocated to other plant controlled documents. The movement of other loads (i.e.,

loads other than fuel assemblies) over irradiated fuel assemblies is administrative 1y controlled based on available analysis for the individual load.

The load analysis and crane operations are described in the UFSAR.

Improved TS 3.7.7 will only be applicable during movement of irradiated fuel assemblies in the associated fuel storage pools.

7.

Existing TS 3.9.9, " Water Level - Spent Fuel Storage and Upper Containment Fuel Pools," is applicable whenever irradiated fuel assemblies are being stored in the spent fuel storage or upper

! A 1

'112 containment fuel pools. The existing TS states that at.least 23 feet of water shall be maintained over the top of irradiated fuel assemblies stored in the these pools.

Improved TS 3.7.7, which will still require 23 feet of water over the irradiated fuel assemblies, will be less j

restrictive because it will only be applicable during' movement (not j

simply storage) of irradiated fuel assemblies in the associated fuel L

storage pools. This change'is acceptable to the staff because the only time high water level is needed is to mitigate the consequences of the design basis fuel handling accident. The bounding design basis fuel handling accident over the spent fuel storage pool assumes an irradiated i

fuel assembly is dropped onto an array of irradiated fuel assemblies seated within the pool racks. Because the licensee's analysis assumes 60% of the damaged fuel rods are in the dropped assembly, the i

consequences for an event other than a dropped fuel assembly would be l

significantly reduced. Therefore, the requirements specified in the existing TS are relocated to the UFSAR or TS Bases and will be controlled in accordance with 10 CFR 50.59.

j 8.

The requirements of existing TS 3.7.4, " Snubbers,"' 3.7.5, " Sealed Source j-Contamination," 3.7.8, " Area Temperature Monitoring," 3.7.9, " Spent' Fuel j

Storage Pool' Temperature" and 3.7.10 " Flood Protection" do not affect i

reactor operation, do not state parameters that are. initial condition-1 assumptions for a DBA or transient, do not state.a significant abnormal i

degradation of the reactor coolant pressure boundary, and do not provide l

for any mitigation of.a design basis event. Therefore,~the requirements specified in these existing TS do not satisfy the NRC Policy Statement screening criteria and have been relocated to the UFSAR or TS Bases and will be controlled according to 10 CFR 50.59.

)

9.

Existing SR 4.11.2.7.1 specifies requirements to monitor the release j

rates of noble gases near the outlet of the main condenser air ejectors in accordance with the ODCM, which is in existing TS 6.14. The licensee l

will be required to use the ODCM by the Administrative Controls in 1

improved TS Section 5.5.1.

The OCDM contains a description of the j

radioactive effluents controls program as improved TS 5.5.4. -Therefore, i

requirements in existing SR 4.11.2.7 are relocated to the UFSAR or TS Bases and will be controlled in accordance with 10 CFR 50.59.

I

10. The licensee proposes moving to the improved TS Bases, existing TS i

3.7.1.3 details of tower basin system design and purpose and their effect i

on operability of the ultimate heat sink (UHS) in Modes 1, 2, and 3.

.1 These details are currently in existing TS 3.7.1.3.

The UHS basin is a l

support system for the required high-pressure core spray service water system (HPCS SWS). Therefore, the requirement for.HPCS SWS operability requires the UHS basin be capable of performing its required support function. The UFSAR also describes the design features and operations for the HPCS SWS and the UHS. The provisions in improved TS Chapter 5 l

will control changes to the Bases.

l

11. The licensee proposes moving to the improved TS Bases, existing TS 3.9.9 l

details of the methods for meeting existing TS 3.9.9 requirements to stop moving irradiated fuel assemblies if the water level limits of the LCO i

_. -. _ =

r

...r e,_.,,..

c.,m_r...-_m.y

..r..

1 i

113 i

. are not met by placing the fuel assemblies in a safe position. This detailed information is descriptive and need not be in the improved TS.

Improved TS Chapter 5 will control changes to the Bases.

i The above relocated requirements relating to plant systems are not required to j

be in the TS under 10 CFR 50.36, and are not required to obviate the j

possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety. Further, they do not fall within any of the four criteria set forth in the Commission's Final Policy Statement, discussed in the Introduction above.

In addition, the staff finds that sufficient regulatory controls exist under 10 CFR 50.59. Accordingly, the i

staff has concluded that these requirements may be relocated from the TS to

)

i the licensee's TS Bases, or the UFSAR, as applicable.

4 C.

More Restrictive Requirements.

By electing to implement the N EG 1434, Section 3.7 Specifications, the licensee proposed the following more restrictive conditions than the existing 2

i TS require. The more significant conditions are the following:

1.

The amount of increase in existing TS 4.11.2.7.2-(improved TS SR 3.7.5.1) is changed from " greater than 50%" to " greater than or equal to 50%."

i This change is more restrictive since, technically, it increases the range of releases to be considered. However, no additional performances 1

of the surveillance would be expected since the increase is j

insignificant. Additionally, if the source of radioactivity is not isolated, or if the mechanism for the release is not removed from service, the plant must be brought to a cold shutdown to eliminate the source.

j' 2.

Improved TS 3.7.4 is added which specifies requirements for operability of the control room air conditioning system. This LCO is necessary to assure the habitability of the control room in a post design basis accident environment.

4 4

i The staff has reviewed the above more restrictive requirements and concludes i

they result in an enhancement of the existing TS.

Thus, these changes are j

acceptable.

4 D.

Less Restrictive Requirements.

By electing to implement the NUREG-1434, Section 3.7, Specifications, the i

licensee proposed several less restrictive conditions than the existing TS i

allows. These conditions are as follows:

1.

In implementing improved TS SR 3.7.1.4, SR 3.7.2.2, and SR 3.7.3.3,-the i

phrase " actual or simulated," in reference to the automatic initiation signal, is added to existing TS 4.7.1.1.b, existing TS 4.7.1.2.b, and a

existing TS 4.7.2.d.2 for verifying that each subsystem actuates on an j

automatic initiation signal. This allows satisfactory automatic system i

initiations for other than surveillance purposes to be used to fulfill

)

the surveillance requirements. Since the system cannot discriminate Y

l 4

114 between " actual" or " simulated," either case adequately shows

^

operability.

2.

The' requirement in existing TS 4.7.2 a. to perform system testing "on a Staggered Test Basis" is deleted. Since the-frequency was not affected (both the existing TS and improved TS require monthly testing for_each subsystem) and scheduling is not a safety concern as long as both subsystems are not tested simultaneously, this is deleted with no impact t

on safety.

{

3.

The applicability of existing TS 3.7.2 is revised to exclude Mode 4 and 5 l

if no activities are being conducted which may lead to a need for control i

j room ventilation system operation in the high radiation mode. The probability and consequences of a design basis accident are significantly reduced due to the temperature and pressure limitations in these modes.

j However, some activities increase the probability of some accidents and these are retained in the applicability.

t 4

4.

Improved TS 3.7.1 includes an additional condition 'for a single f

l inoperable fan in one UHS cooling tower with a 7-day time for repair.

The inoperability of a single fan would not make'the standby service water system completely unavailable to perform its safety function but j

would reduce the heat removal capability of the system. The inoperable j

fan would not affect the redundant subsystem. During the 7-day completion time for restoring system operability, the probability of a design basis accident is low and the system is degraded; however, heat j

removal capability is available using the operable subsystem.

5.

The licensee proposes new Conditions B and C for improved TS 3.7.1 to l

clarify the interaction between the SWS and the UHS basin and cooling tower fans. This change accounts for the design of the SWS and UHS, l

which must be treated as separate systems with separate applicable I

conditions. Otherwise, the UHS basin would have to be considered a i

support system, which, if inoperable, would require both SWS subsystems to be declared inoperable and result in an immediate shutdown. The shutdown would be unnecessary because the UHS basin would become j

inoperable only if the water level fell, which would occur slowly and i

only slightly reduce the capability of the UHS to perform its analyzed l

function. A completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for this condition is sufficient i

time to correct the situation and prevent the unnecessary shutdown.

j Separate conditions and ras allow the SWS and the UHS basin and cooling tower fans to be treated independently.

Improved TS 3.2.1, Condition E, 1

changes the first condition to include the new conditions not met and i

adds a fourth condition "QB Two UHS cooling towers with one or more j

cooling tower fans inoperable."

The above less restrictive requirements have been reviewed by the staff and j

have been found acceptable because they do not present a significant safety question in the operation of the plant. The TS requirements that remain are 4

consistent with current licensing practices, operating experience, and plant i

accident and transient analyses, and provide reasonable assurance that the public health and safety will be protected.

t 4

)

1

.=

jo C

115 3.8 Electrical Power Systems A.

Significant Administrative Changes.

In accordance with the guidance in the Final Policy Statement, the licensee has proposed administrative changes to the existing technical specification (TS) to bring them into conformance with the improved TS. These changes are i

as follows 1.

Existing TS 3.8.1.1 ' indirectly require actions leading to hot and cold shutdown with three Alternating Current (AC) sources inoperalle (existing TS 3.0.3).

The improved TS 3.8.1 Condition G requires immediate entry 1

into improved TS limiting condition for operation (LCO) 3.0.3 if three or more AC sources are inoperable. Without improved TS Condition G.the format of NUREG-1434 would allow multiple conditions for inoperable AC sources to be simultaneously entered. As a result, three required AC sources could be inoperable, ras taken in accordance with the specification, and yet improved TS LCO 3.0.3 entry would.still not be required. To preserve the existing intent for improved TS LCO 3.0.3 entry, the improved TS includes this new condition.

Since the improved TS changes effectively retain the existing TS 3.8.1.1 required actions, the changes are administrative and are acceptable..

2.

Existing TS 3.8.3.1 indirectly require. actions leading to shutdown

{

(existing TS 3.0.3) with distribution subsystems in both Division 1.and 2 1

inoperable.

Improved TS 3.8.7 RA E.1 requires immediate entry into LCO 3.0.3 if two or more divisions have inoperable distribution subsystems.

Without the RA, the improved TS format would allow the loss of redundant distribution subsystems in separate divisions, which would result in multiple entry into conditions with ras being taken in.accordance with j

the Specification, and yet avoid the immediate shutdown requirement included in the existing TS. The improved TS RA conservatively assures i

that if multiple distribution subsystems become inoperable resulting in 1

the inability to adequately respond to an accident, an immediate shutdown will be required. Since the improved TS changes effectively retain the existing TS requirements, the changes are administrative, and are acceptable.

3.

The existing TS 3.8.1.1 separates actions for Divisions 1 and 2, from actions for Division 3 (HPCS) sources and distribution systems. The existing TS does not require Division 3 sources or the associated DG to be to be operable when the HPCS system is inoperable.

The format of the improved TS combines the requirements for all DGs, requiring the three DGs connected to the three division load groups to be operable during Modes 1, 2, and 3.

However, the Applicability Note for the improved TS 3.8.1 states that the HPCS DG is not required to be operable when the 1

HPCS is inoperable. The note allows the licensee to continue applying the conditions and ras for other inoperable AC sources without addressing i

the HPCS DG inoperability, which is addressed in improved TS 3.5.1.

The improved TS Applicability exception is consistent with the existing TS L

l4 e

i i

i 116 4

3.8.1.1, which separates actions for Divisions 1 and 2 from actions for I

Division 3.

The improved TS retain the changes and are therefore j

acceptable.

I 4.

The existing TS does not prohibit the separate entry into the conditions j

and ras of both the support system and supported system if a required i

l support system becomes inoperable. However, the improved TS adds LCO 3.0.6, which excludes the entry into the conditions and ras for a i

supported system when it becomes inoperable unless otherwise specifically stated in the LCO. The AC sources are considered a support system to the' i

distribution system in the improved TS.

If AC sources are inoperable such that a distribution subsystem were inoperable, the improved TS LCO i

3.0.6 would allow taking only the AC source actions and not taking the AC

{

1 l

Distribution system action. Specific direction to take appropriate actions for the distribution system is required because the AC source actions are not sufficiently conservative in this event. Condition D-RA

{

e Note in improved TS 3.8.1 requires the entry into the applicable

{

conditions of TS 3.8.7, " Distribution Systems - Operating," when the concurrent inoperability of one AC offsite circuit and one DG results j

eliminates AC power for a required division. The change is administrative and therefore acceptable because the improved TS does not change the existing TS allowed actions and only establishes a procedure l

for current practice.

i 5.

The existing TS 4.8.1.1.2.d.16 includes an 18-month SR to verify that DG i

lockout features prevent the DG from starting or trip'only when required.

l The lockout features include no control power, low starting air pressure, and an energized stop-solenoid. The improved TS does not-include this SR, since the DG is started monthly and the failure of these devices to 4

operate correctly would prevent the start. The improved TS need not include the existing TS 18-month surveillance requirement because the existing TS requirement is done only on a once a month basis' The e

i improved TS changes are administrative and are therefore acceptable l

l because they effectively retain the existing TS 3.8.1.1-required actions.

1

]

6.

The licensee proposes adding a note to the following improved TS to state that LCO 3.0.3 does not apply: "A.C. Sources - Shutdown" (3.8.2), "D.C.

Sources - Shutdown" (3.8.3) and " Distribution Systems - Shutdown" (3.8.8). LCO 3.0.3 requires the licensee toshut down the reactor if the actions of the LCO are not done. Operations involving moving irradiated fuel assemblies while the plant is in Mode 1, 2, or 3 are independent of

~

j reactor operations. Therefore, inability to suspend movement of i

irradiated fuel assemblies is not sufficient reason to require a reactor j

shutdown, and the improved TS clarify the implicit requirements of the l

existing TS.

i 7.

The limitations in existing TS 3/4.8 to do 18-month surveillances "during shutdown" are more clearly presented in the improved TS surveillances.

j Each 18-month SR contains a note limiting the performance in certain i

modes. While these limitations vary among SRs, each is consistent with r

NUREG-1434, which defines the intent of "during shutdown" for the individual SR, consistent with the guidance of Generic Letter 91-04. The I

i j

1

- - - J

jo 117 note clearly presents the allowance of the current practice of taking credit for unplanned events if the necessary data are obtained. This i

change involves only the presentation of current practice and is

)

administrative.

8.

The improved TS do not include cross references such as the reference in

{

existing TS 3.8.1 to "take the ACTION required by Specification 3.5.1."

)

This type of reference has no safety significance and thus can be removed as an administrative preference.

)

9.

The requirement in existing TS 4.1.1.2.a.4 to obtain a specified minimum i

speed (rpm) is also statM in the requirement to obtain the specified frequency (Hz) range sir.ce the minimum frequency range is at the same 4

i point. Therefore the minimum speed limit is not separately included in improved TS SR 3.5.1.2.

The change is administrative because the requirement for a specified frequency range is retained.

i i

10. The licensee proposes adding two notes to more clearly present the current methodology for doing improved TS SR 3.8.1.3.

This surveillance

{

demonstrates that the DGs are capable of synchronizing and accepting loads greater than or equal to the maximum expected accident loads. A i

minimum run time of 60 minutes, or until engine temperatures stabilize is required while minimizing the time that the DG is connected to the offsite source. Note 3 states that this surveillance shall be done on 1

i only one DG at a time to avoid common cause failures that might result i

j from offsite circuit.or grid perturbations. Note 4 stipulates a i

prerequisite requirement for doing this SR. A successful DG start must j

precede this test to credit satisfactory performance. This change is i

j administrative because it affects only the TS presentation of current i

practice.

j u

11. The 92-day frequency of improved TS SR 3.8.6.3 does not require performance after a battery is discharged or overcharged. However, the battery overcharge frequency requirement is included in improved TS SR 3.8.6.2, which includes a check of the specific gravity, which must be i

corrected for electrolyte temperature. Therefore, the temperature must still be obtained for the verification of operability of the battery.

This change in TS presentation is administrative.

12. The automatic load sequencers specified in existing TS 3.8.3.1 and j

3.8.3.2 operate to give the appropriate integrated response for starting DGs, offsite circuits, and individual components.

Various failure i

]

mechanisms can be postulated for " automatic sequencers." Each failure l

requires the operator to ascertain the effect of the failure on each of i

the functions, and determine the effect on the operability of each function. Appropriate actions are provided in improved TS 3.8.2 for an inoperable DG or circuit or in the action statements for an inoperable individual system caused by automatic load sequencer failures.

Therefore, no separate LC0 item is proposed.

i i

i t

4

.yy c

g

a0 f

118 J

l The above' changes result in the same limits as the current _ requirements, or they represent an enhanced presentation of~the existing TS intent.

Accordingly, the improved TS changes are purely administrative and they are acceptable.

l B.

Relocated Requirements In accordance with'the guidance in NUREG-1434,-the licensee has proposed to j

relocate all or portions of the following-existing TS within the improved TS:

j Existina TS Ilt.].g 3/4.8.1.1 AC Sources - Operating 3/4.8.2.1 DC Sources - Operating i

3/4.8.4.3 Reactor Protection System Electric Power i

Monitoring a

The more significant changes resulting from relocated items are as follows:

i 1.

The technical requirements for Diesel Fuel 011 and the testing thereof, l

presently included in existing TS 4.8.1.1.2 are being moved to improved TS 5.5.9, " Diesel Fuel Oil Testing Program." The technical requirements and testing of diesel fuel oil are a quality concern and have no

)

immediate direct effect.on the operability of Diesel Generator (DGs).

l 2.

Existing TS 3.8.4.3, " Reactor Protection System Electric Power j

Monitoring," has been moved to improved TS 3.3.8.2.

The existing TS requirements are more appropriately instrumentation requirements and therefore are included in the " Instrumentation" section 3.3 of the 4

i improved TS.

o l

4 3.

The improved TS present the battery cell parameter limits from existing i

TS 3.8.2.1 in a separate LC0 (improved TS LCO 3.8.6).

The references to hardware components (battery and charger) remain _in improved TS LCO 3.8.4 j

for DC sources.

l The above changes are considered administrative, since the existing TS requirements are only relocated within the TS, and the changes are therefore i

acceptable.

l In accordance with the guidance in the Final Policy Statement, the licensee has proposed to relocate or reorganize all or portions of the following existing TS to other licensee-controlled documents:

Existina TS Title 3/4.8.1.1 AC Sources - Ope' rating i

3/4.8.1.2 AC Sources - Shutdown 3/4.8.2.1 DC Sources - Operating 3/4.8.3.1.2 Distribution - Shutdown 3/4.8.4.1 Containment Penetration Conductor Overcurrent Protective Devices 3/4.8.4.2 Motor Operated Valves Thermal Overload i

Protection i

i

ja

?-

l 119 i

The more significant changes resulting from relocated items are as follows:

1.

Existing TS 3.8.4.1 describes primary containment penetration conductor overcurrrent protective devices. These devices provide protection for the circuit conductors against damage or failure due to overcurrent i

4 i

heating effects, but are not relied upon in any design basis accident or transient.

Further, the evaluation summarized in General Electric's study of TS, NED0-31466,. determined the loss of these protective devices to be a non-significant risk contributor to core damage frequency and 3

offsite release. Therefore, the requirements specified for-this function 1

have been relocated to the UFSAR or TS Bases and will be controlled f

in accordance with 10 CFR 50.59.

l 3

2.

Existing TS 3.8.4.2 describes motor operated valve thermal overload

}

protective devices. These devices protect the equipment from potential damage to maintain the capability of the equipment, but are not relied 4

j upon in the primary success path to mitigate a design basis accident or i

transient. Therefore, the requirements specified for this function have

]

been relocated to the UFSAR and TS Bases and will be controlled in y

accordance with 10 CFR 50.59.

i j

3.

Existing TS 4.8.1.1.3 includes reporting requirements for all diesel i

generator test failures. The requirement for a Special Report following diesel generator failures' the associated details of the report, and when

{

it should be submitted can be adequately controlled by relocating the i

i requirement to the UFSAR or TS Bases and controlled in accordance with 10 I

i CFR 50.59. Deletion of this requirement from the existing TS is consistent with changes outlined in NRC's Generic Letter 94-01, " Removal i

of Accelerated Testing and Special Reporting Requirements for Emergency l

Diesel Generators," issued on August 30, 1994, i

I 4.

Existing TS 3.8.1.2, Action a., which suspends crane operation following j

a loss of all required AC power sources, has been relocated to the UFSAR.

Crane operation is not directly affected by loss of safety related power sources and is not required to prevent or mitigate the consequences of a DBA or fuel handling accident.

5.

The surveillance procedures in existing TS 4.1.1.2.a.4.a) through d) i dictate the starting methods and signals for DG fast start capability.

j This SR helps to ensure the standby electrical power supply is available 3

to mitigate DBAs and transients and maintain the unit in a safe shutdown j

condition. The licensee proposes removing available options currently 1

described in the TS.

Procedural controls and the revised description of i

these details are sufficient guidance.

j 6.

The surveillance procedures in existing TS 4.1.1.2.a.6 include procedural controls on DG standby alignment. The definition of DG operability and the associated action requirements are sufficient to ensure the DG remains aligned to provide standby power. This surveillance is not i

discussed in the improved TS. Relocation of these details from the TS to the UFSAR will not affect DG operability.

)

l k

i

120 i

7.

The surveillance requirement in existing TS 4.1.1.2.d.1 ensures that DG f

inspections recommended by the manufacturer are done.

This surveillance is not specifically described in the improved TS. Relocation of these l-details to the UFSAR will not affect DG operability.--

j 8.

The surveillance requirement in existing TS 4.1.1.2.d.2 specifies the kW 2

requirements for the single largest load allowed for the DG load reject i

test. The value of the load and the name of the component are. stated in

}

the plant design in the UFSAR, and changes to them are controlled by 10 CFR 50.59. The licensee proposes relocating these references and the i

load value for the automatically connected loads to the UFSAR..

Therafore, the TS need not refer to them to adequately control the-j requirement.

}

9.

The licensee proposes moving details on system design for auto-connecting i

emergency loads from the existing TS SR 4.8.1.1.2.d.6.b)2) to plant procedures. The requirement to verify the connection and power supply of permanent and automatically connected loads is intended to satisfactorily 4

j show the relationship of these loads to the loading logic for loading i

onto offsite power.

In certain circumstances, many of these loads cannot i

be connected or loaded without undue hardship or possible undesired

]

operation. For instance, ECCS injection valves are not desired to be stroked open, high-pressure injection systems are not capable of being operated at full flow, or RHR systems removing decay heat are not desired j

to be realigned to the ECCS mode of operation. Instead of demonstrating i

the connection and loading of these loads,' the licensee may do testing that adequately shows the capability of the offsite power system to do these functions. The design features and system operation are also described in the UFSAR. Changes to these procedures will be controlled by the provisions of 10 CFR 50.59.

+

10. The details of system design and the purpose of DC power sources in existing TS 3.8.2.1, and 3.8.2.2, and for onsite power distribution systems in existing TS 3.8.3.1, and 3.8.3.2 have been relocated 1o the TS i

Bases and UFSAR. The design features and system operation are also i

described in the UFSAR. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Program in Chapter 5 of the TS.

a i

11. The existing TS 4.8.1.1.2.d.16.a) through 1) includes a list of specific i

bypassed DG trips, and DG lockouts to be confirmed as bypassed or not present, when required to be bypassed or not present to verify that DG lockout features prevent DG starting or trip only when required. The i

list is removed from the improved TS. The requirement for all i

unnecessary trips to be bypassed (or not present) adequately controls the 1

intent of these requirements. The specific lists are adequately j

controlled by plant procedures and the 10 CFR 50.59 revision process.

12. The licensee is required to do postmaintenance testing to demonstrate a 1

system or component is operable whenever the operability of the system or component has been affected by repair,. maintenance, or replacement of a component. Therefore, the licensee proposes deleting explicit postmaintenance requirements from existing TS SR 4.1.1.2.e.

The improved i

i i

4 I

\\

~

121 i

TS Bases for SR 3.0.1'do not allow the licensee to place the plant in the i

applicable modes without doing postmaintenance testing.

13. The licensee proposes modifying the TS to remove the description of the use of sodium hypochlorite for removing accumulated sediment and for j

cleaning fuel oil storage tanks (existing TS SR 4.1.1.2.e.1).

The licensee also proposes removing the description ;in existing TS,SR

]

4.1.1.2.e.2 of pressure tests of fuel oil systems designated in the ASME Soffer and Pressure Vessel Code. The details are adequately controlled j

by plant procedures and the 10 CFR 50.59 revision process.

j i

1 1

14. The licensee proposes moving details of the methods for meeting existing i

TS SR 4.8.3.1.2 to the TS Bases and UFSAR.

TS SR 4.8.3.1.2-describes how to demonstrate the manual test inputs for the load shedding and l

l sequencing panels. The UFSAR describes the design features and system operation that dictate the methods described therein.

Changes to the'TS j

Bases will be controlled by the proposed Bases Control Program in Chapter i

5 of the Technical Specifications.

i

l 1

15. The licensee proposes moving' to plant procedures the restriction in i

j existing TS SR 4.8.1.1.2.e that the 10-year simultaneous start of all three DGs be performed "during shutdown." This detail.is adequately i

j controlled by plant procedures and the 10 CFR 50.59' revision process.

i

16. Existing TS SR 4.8.1.1.2.d.9 specifies the HPCS diesel' generator voltages 3

and frequency for the required 24-hour test run.

Limits of the frequency l

and voltage during the-24 hour run are unnecessary because this test. is performed with the DG connected in parallel to offsite power, and the j

power factor which is.to be maintained is specified. These DG operating parameters are sufficiently verified by other surveillances, and are 1

4

{

sufficiently controlled by plant procedures, offsite power (grid)

'I requirements, and the new power factor requirement. The details of J

voltage and frequency can be adequately controlled by plant procedures j

and the 10 CFR 50.59 revision process.

I

17. Existing TS SR 4.8.2.1.c.2 is a test of the average electrolyte-l temperature. The IEEE-450 industry standard, from which the TS tests is j

derived, requires only that " representative" cells be measured for temperature; every sixth cell is only suggested. Therefore, the licensee proposes moving the detail on the plant-specific determination of j

" representative" to the Bases.

18.

Performance of the " service test" as the controlling requirement in improved TS 3.8.4.7 is adequate to convey that the duration for the i

battery capacity test must be' consistent with the plant-specific licensed service duration. The licensee proposes moving the reference to times of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for Divisions 1 and 2 and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for Division 3 from existing TS

}

SR 4.8.2.1.d to the Bases and plant procedures.

i 1

19.

Existing TS SR 4.8.2.1.f requires an annual battery discharge test for j

those batteries that show " degradation." Degradation is defined as a

[

fractional loss of the manufacturer's rated capacity as compared to the l

l i

I 122 i

l average from previous performance tests. The licensee proposes moving the definition of " degradation" to the Bases and plant procedures.

i _

The above relocated requirements relating to electrical power systems are not j

required to be in the TS under 10 CFR 50.36, and are not required _ to obviate the possibility of an abnormal situation or event giving rise to an immediate j

threat to the public health and safety. Further, they do not fall within any i

of the four criteria set forth in the Commission's Final Policy Statement, discussed in the Introduction above.

In addition, the staff finds that sufficient regulatory controls exist under 10 CFR 50.59. Accordingly, the s

staff has concluded that these requirements may be relocated from the TS to l

the licensee's TS Bases or in the UFSAR, as applicable.

C. More Restrictive Requirements i

By electing to implement the NUREG-1434 Section 3.8 Specifications, the licensee has adopted a number of more restrictive conditions than are required i

j by the existing TS. The more significant conditions are the following-1 4

1.

Improved TS 3.8.3, " Diesel Fuel Oil, Lube Oil, and Starting Air" includes 3

l new TS requirements for DG lube oil. The requirements are necessary to ensure that a sufficient lube oil inventory is maintained to provide t

proper lubrication for operation of the associated DG under all required loading conditions.

l 2.

The existing TS has no requirement to declare redundant features (i.e.,

system, subsystem, component, etc.) inoperable upon the concurrent inoperability of a required redundant feature and the two required AC 3

offsite circuits.

Improved TS 3.8.1 RA C.1 requires that redundant features be declared inoperable upon the concurrent inoperability of a i

required redundant feature and the two required AC offsite circuits. The improved TS action will reduce the vulnerability.to an unexpected loss of 4

j function caused by the failure of the DG supplying the operable function during the time (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) allowed by existing TS 3.8.1.1 (improved TS 3.8.1 RA C.2) to restore one of two inoperable offsite AC circuits to operable status.

a 4

l 3.

Existing TS 3.8.1.1 does not include any actions to address an inoperable offsite AC circuit that leaves the two AC divisions with no offsite i

power.

Improved TS 3.8.1 RA A.2 requires the restoration of an inoperable offsite AC circuit to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the j

inoperability leaves two Class IE electrical divisions with no offsite i

power. The improved TS restriction is required because three AC divisions are required and only two AC offsite circuits are required.

Thus, a single inoperable circuit could leave multiple divisions without power.

In this situation, the more restrictive outage time in the l

improved TS is appropriate.

1 4.

Existing TS 3.8.1.1 requires a confirmation of the status of other required systems supplied by the operable DG with either the Division 1 or 2 DG inoperable.

Improved TS 3.8.1 RA B.2 requires that when any DG

[

(Division 1, 2 or 3) becomes inoperable, the required feature (s)

a p

l 1

I 123 supported by the inoperable DG be declared inoperable when the redundant s

features are inoperable. With a DG inoperable, certain General Design l

l Criteria (GDC) 17 required events may not be capable of being mitigated i

if other required systems are also inoperable. To ensure events can be l

mitigated, the existing TS require a confirmation of the status of other required systems supplied by the operable DG with either the Division 1 1

.l or 2 DG inoperable. However, certain combinations of HPCS inoperability 1

along with the inoperability of other emergency core cooling systems I

(automatic depressurization system and/or low pressure core spray) are also in this category. To ensure that all events can be mitigated with the loss of offsite AC power, the improved TS RA requires that when any l

DG becomes inoperable, the required feature (s) supported by the.

i inoperable DG be declared inoperable when redundant required features are inoperable. This more restrictive requirement adds operational j

conservatisms not included in the existing TS.

)

5.

Existing TS 3.8.1.1 allows the Division 3 DG and either the Division 1 or

\\

f Division 2 DGs to be concurrently inoperable for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Improved TS 3.8.1 RA E.1 limits any concurrent inoperability of both the l

j Division 3 DG, and either the Division 1 or 2 DG up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> + --If both DGs are still inoperable after the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, the HPCS may be 4

j declared inoperable such that the remainder of the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, as allowed by improved TS 3.8.1 RA B.4 for one DG inoperable, could be.used to restore the inoperable Division 1 or Division 2 DG tc operable status, i

This improved TS change adds operational conservatisms.iot included in the existing TS, since it places a more restrictive time prior to having to declare the HPCS inoperable than the currently allowed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

l 6.

Existing TS 4.8.1.1.2 does not require the licensee to check for accumulated water in the diesel fuel storage tanks and remove any that it i

finds. The licensee proposes improved TS SR 3.8.3.5, which includes this j

requirement with a 92-day frequency to ensure that water will not degrade j

the performance of the DG, i

7.

Existing TS 4.8.1.1.2.e DG simultaneous start test only requires the-i verification that the engine speed is above minimum Revolution Per Minute (RPM) limits for each DG.

Improved TS SR 3.8.1.20 adds requirements to verify that the DG output is within minimum / maximum voltage limits.

l These requirements are consistent with other DG start SRs.

l l

8.

Existing TS Table 4.8.1.1.2-1, " Diesel Generator Test Schedule," requires weekly diesel generator testing if previous testing identifies excessive failures.

In order to resume monthly testing under the existing TS, seven consecutive failure-free tests must be performed and the number of failures in the last 20 valid tests is less than or equal to one.

In addition, staff guidance requires that these starts must be " cold starts".

In order for these tests to qualify as " cold starts," improved TS Table 3.8.1-1 requires that at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> elapse between each test. The improved TS added restriction will assure that'the seven consecutive tests are performed with a cold DG.

i e

T n

e N,u-

.,p r

--,,,. -, +.,

-m--,

~e-

1.

j 124 9.

-Existing TS 4.8.1.1.2 does not include any generator power factor

'l requirements in the performance of the DG load rejection and 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> load l

test. A specific power factor has been added to the requirements for

~l improved SR 3.8.1.9 and 3.8.1.10 (load rejection tests) and improved SR 4

3.8.1.14 (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> load test) when the DG is synchronized with offsite power during the test. These more restrictive limitations ensure that i

each DG is conservatively tested at as close to accident conditions as.is reasonably achievable.

10. Existing TS SR 4.8.1.1.2.d.5 tests the DG capability to automatically j

start from standby condition on either an actual or simulated ECCS J

signal.

Improved TS 3.8.1.12 includes an additional acceptance criterion l

to verify that required emergency loads are automatically connected to' the offsite power system. This surveillance is necessary for the system' to do its intended function.

1 4

a

11.. The existing TS LCO requires that one offsite AC circuit be operable l

during shutdown.

Improved TS 3.8.2 requires that one qualified circuit i

between the offsite transmission network and the onsite Class IE AC-j electrical distribution subsystem (s) required by TS 3.8.10, " Distribution 1

Systems-Shutdown". The existing TS LCO requirement that one offsite i

circuit is able to be operable during shutdown conditions is not specific as to what that circuit must be powering, or even if it must be capable I

l of supplying power to the required distribution system (s). The improved TS provides assurance that the required offsite circuit is supplying the i

required distribution system (s). The improved TS result in a more-restrictive change because it enforces a level of TS control for matters j

which currently must be enforced via administrative procedures.

12. Existing TS 3.8.1.2 requires that the Division 1 or 2 DG be operable during shutdown.

Improved TS 3.8.2 requires that one DG capable of i

supplying the one division of the Division 1 or 2 onsite Class IE' AC electrical distribution s.ubsystem(s) required by LCO 3.8.10,

" Distribution Systems-Shutdown," be operable.. The existing TS LCO i

requirement is not specific as to what distribution system (s) the one l

required DG must be capable of powering. The improved TS is a more restrictive change because it enforces a level of TS control for matters which currently must be enforced.via administrative procedures.

)

13.

Existing TS 3.8.1.2 does not. require the restoration of inoperable

{

required AC sources to operable status when the reactor vessel cavity is j

flooded during shutdown. The improved TS eliminates this allowance.

t l

14.

Existing TS 4.8.2.1.b requires that within seven days after a significant battery overcharge or discharge the battery cell parameters be verified d

to meet the Category B limit.

Improved TS 3.8.6.2 reduces the allowed j

time to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following an overcharge. A battery overcharge, which brings the battery terminal voltage to greater than 150 volts, can have a significant effect on the batteries' capability and confirmation of continued battery operability within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is a more restrictive change.

For a battery discharge below 110 volts, the battery would be i"

i l

125 l

inoperable and the category B limits would have to be restored prior to j

declaring the battery operable.

j 15.

Improved TS Table 3.8.6-1, "Sattery Cell Parameters Requirements,"

deletes the existing TS Table 4.8.2.1-1, " Battery Surveillance s

i Requirements," option to correct the battery float voltage (Category B

{

i-limit) measurement based on the cell electrolyte temperature. The i

removal of this optional allowance puts in place a more restrictive l

l requirement in meeting Category B voltage limits.

16. When the battery is on float charge, existing TS Table 4.8.7.1-1,

]

footnote (c), allows a battery charging current of less than 2 amps to be 2

used indefinitely in lieu of meeting the " pilot cell" and " allowable.

j value for each connected. cell" specific gravity limits. With a battery i

i on float charge, improved TS Table 3.8.6-1, Note (c) allows a charging.

. current of less than 2 amps to be used in lieu of meeting all specific-i '

gravity limits for a maximum of seven days following a battery recharge.

The use of charging current in lieu of specific gravity is allowed since l

the charging current provides a satisfactory indication of battery cell condition for a limited time after a battery recharge when specific gravity readings may be unreliable. By limiting the time that the i

charging current may be used in lieu of specific gravity, the use of the.

charging current is ensured to be sufficiently conservative.

j 17.

Improved TS 3.8.6 RA A.1 requires that the battery pilot cell electrolyte

]

level and float voltage be verified within Table 3.8.6-1 Category C L

limits when one or more batteries with one or more battery cell i

parameters is not within Table 3.8.6-1 Category A or B limits. This i

requirement is not included in the existing TS.

In addition, improved TS

)

i RA A.2 includes a new requirement that all battery cell parameters must be verified to be wi. thin Category C limits within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and then every i

seven days until the batteries are returned to Category A and B limits.

The improved TS provides assurance that the batteries have the capability d-to perform their safety function during the increased time (31 days) that the batteries are allowed to be outside the Category A and B limits.-

i 18.

Existing TS 3.8.2.2 requires that either Division 1 or-2, and Division 3 j

j and 4 (when the HPCS is required) Direct Current (DC) sources shall be i

l operable.

Improved TS 3.8.5 additionally requires the second Division 1 l

or 2 battery or battery charger to be operable when the Division 1 and 2 distribution systems are required by improved TS 3.8.10, " Distribution 4

l Systems-Shutdown." The existing TS LCO requirement for Division 1 or Division 2, and Division 3 and Division 4-(when the HPCS is required) DC i

sources to be operable is not specific as to what the required system (s)

I l

must be powering. The improved TS is a more restrictive requirement i

because it conservatively assures the needed DC sources are operable.

I f

3

19. Existing TS 3.8.2.2 has no actions to restore DC sources to operable status after the suspension of all Core Alterations, irradiated fuel handling, and operations with the OPDRVs.

In the event that necessary DC shutdown sources are not operable, plant conditions are conservatively j

restricted by suspending all Core Alterations, irradiated fuel handling, J

i-i

..... ~..

., _,..., _.., _. ~.,

j*

1 J

j 126 and OPDRVS. However, continued operation without the necessary DC j-sources should not be considered prudent.

Improved TS 3.8.5 RA A.2.4 j

requires that actions be taken immediately to restore the required DC electrical power subsystems and to continue' this action until restoration y

is accomplished in order to provide the necessary DC electrical power to the plant safety systems. The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required DC electrical power 4

subsystems should be completed as quickly as possible in order. to 1

minimize the time during which the plant safety systems may be without l

power.

i i

20. Existing TS 3.8.3.1, "Onsite Power Distribution Systems," describes the i

minimum power distribution system divisions that must be operable. The I

associated Action statement permits separate allowed outage times of-i eight hours for an inoperable AC distribution system and two hours for an

{

inoperable DC distribution system. However, there is no link between the

~

allowed outage times for AC and DC systems.

This has created the possibility that the AC and DC distribution systems could fail consecutively such that an indefinite period of time could occur when the l

LCO requirements are not fully satisfied. The existing TS considers the i

inoperabilities of AC, uninterruptable AC bus,. and/or DC distribution i

subsystems ' separately and does not consider the cumulative or contiguous' 3

effect of the inoperable subsystems ~

This.can occur even though all ras i

for the individual distribution subsystems are completed within their l

required completion time.

Improved TS 3.8.9 includes new limitations for j

.the total time allowed for required AC and DC distribution subsystems to j

be inoperable. The staff has concluded that a limit of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> from 4

discovery of failure to meet the LCO establishes a maximum time allowable for any combination of required distribution subsystems to be inoperable during any single consecutive occurrence of failing to meet the LCO.

i i

Therefore, to place an appropriate restriction on any such situation, the improved TS includes a 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> limit on the time that the LCO is not j

fully satisfied.

4 l

21. The licensee proposes improved TS 3.8.8, in which it has modified LCO j

requirements from existing TS 3.8.3.2, " Distribution - Shutdown." The i

improved TS LCO states that "necessary portions of the Division 1 3

Division 2, and Division 3 AC and DC electrical power distribution j

subsystems shall be operable to support equipment required to be operable." The existing TS LC0 requirement for Division 1 or Division 2 (and Division 3 when the HPCS is required) to be operable does not state what the single system must be powering. The more restrictive requirement in the improved TS conservatively ensures the needed distribution subsystems are operable, even if this results in the 2

j distribution systems for both Divisions 1 and 2 being operable.

1 22.

Improved TS 3.8.8 RA A.2.5 states that when one or more required i

electrical distribution systems is inoperable, the associated required j

shutdown cooling subsystem (s) must be declared inoperable and not in operation. This improved TS RA, which is not included in the existing TS, will ensure that appropriate consideration is applied to shutdown I

d 4

l

.~.

..a

127 cooling systems that are without power.

Since the improved TS

" Distribution System-Shutdown" actions may not be sufficiently conservative in the event that a distribution system is inoperable, specific direction to take appropriate actions for the affected shutdown cooling systems is required and provided.

23.

Existing TS 3.8.3.2, " Electrical Power Systems, Distribution - Shutdown,"

does not address an allowed outage time for an inoperable AC or DC distribution system. Upon loss of such systems, the associated Action statement requires a suspension of core alterations, handling of irradiated fuel and OPDRVs. However, continued operation without the necessary electrical power distribution subsystem (s) is not prudent. The improved TS 3.8.8 RA A.2.4 requires that actions be taken immediately to restore the required distribution systems and to continue this action until restoration is accomplished in order to provide the necessary electrical power to the plant safety systems. The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required electrical power distribution systems should be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without power.

The staff has reviewed the above more restrictive requirements and concludes that they result in an enhancement to the improved TS, and therefore are-acceptable.

D. Less Restrictive Requirements The licensee in electing to implement NUREG-1434 Section 3.8 Specifications has proposed a number of less restrictive conditions than are allowed by the i

existing TS. The more significant conditions are the following:

1.

The licensee proposes improved TS 3.8.1, which does not contain the action a. requirement from existing TS 3.8.1.1 to do diesel start tests to demonstrate that required DGs are operable under degraded offsite power conditions. The improved TS include the existing TS surveillance testing schedule, which ensures that the operable DGs can do their intended safety functions.

The reliability and availability of the DGs would not be adversely affected solely by the loss of offsite circuit (s), and the DG should not be required to be started if this condition exists.

In some circumstances, the signal detecting an inoperable AC source will l

automatically start the associated DG, which would already be supplying the safety bus. The DG manufacturer recommends that, once the DG starts, the licensee load that DG before returning it to a standby status.

If the DG is tied to the offsite source, severe weather or other off-normal grid conditions can also cause the loss of DG and leave its safety bus without AC power. NRC Information Notice 84-69 warns against operating DGs tied to offsite power when the unit AC sources are abnormally degraded or threatened. When a DG is operated while connected to offsite sources and nonvital loads, failures in this equipment can degrade DG

a t

128 j

reliability.

Further, a demand for DG start is more likely while the DG is connected to the grid and to nonvital loads for this required surveillance because the offsite AC sources have been degraded.

Therefore, DG availability could be reduced by requiring a DG start when the offsite sources are abnormally degraded.

t 2.

Improved TS 3.8.3 reformats some of the existing TS 3.8.1.1 and 4.8.1.1.2 requirements for diesel fuel oil and starting air by providing a separate t

LCO with requirements for each parameter. Diesel fuel oil and starting air are currently presented as attributes of compliance with the DG LCO via their presentation as surveillances in existing TS 4.8.1.1.2.

The parameters (e.g., oil storage tank level, oil particulates, oil properties, and starting air) supporting DG operability contain substantial margin in addition to the limits which would be necessary for DG operability. As an example, the improved TS retains the requirement to maintain starting air above a specified minimum pressure in order to maintain DG operability. However, certain levels of degradation are i

acceptable to extend the allowances for restoration. During the proposed i

extended periods for restoration of these parameters, the DG would still be capable of performing its intended function.

3.

Existing TS 3.8.1.1 limits the time allowed to restore all multiple concurrent inoperable AC sources to operable status to "72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from initial loss." When a second AC source becomes inoperable just prior to the restoration of another AC source to operable status.and close to the expiration of the initial 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, there is very little or no time to effect repair on the second source. This could result in a forced shutdown of the unit. While simultaneous inoperabilities are expected to be rare, it is also expected that any AC source inoperability would be repaired in a reasonable time (less than or equal to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />). Given the minimal risk of an event during the repair of the subsequent inoperability, the likelihood of a satisfactory return of the AC sources to operable status, and the risks involved with introducing plant transients associated with a forced shutdown, the separate 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period to restore each source to operable status is reasonable. Since this rationale can be taken to extreme with continuous multiple overlapping inoperabilities, a maximum restoration time limit restricting this to six j

days (twice the time allowed by a single 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> restoration period) is also proposed. Therefore, improved TS 3.8.1 presents this as an additional RA A.2 with a completion time of six days from discovery of failure to meet the LCO. This does not change improved TS 3.8.1 Condition E which permits a maximum allowed outage time of two hours when both Division 1 and Division 2 diesel generators are inoperable.

In addition, it does not change TS 3.8.1 Condition G which requires entry l

into LCO 3.0.3 when three or more required AC sources are inoperable.

4.

Existing TS 3.8.1.1 requires that if a diesel generator becomes inoperable due to any cause other than preplanned preventive maintenance or testing, the remaining diesel generators must be started to demonstrate operability. The intent of the DG start testing is to confirm that no common-mode failure has rendered more than one DG inoperable. However, situations have occurred when the cause of DG

i h

i l

129 j

failure was clearly not common-mode, yet the existing TS still required j

start tests for the remaining diesel generators. Unnecessary testing of diesel generators has been identified as a leading cause of long-ters DG j

degradation and reduced reliability.

Minimizing DG starts is~ recommended to avoid unnecessary diesel wear, thereby enhancing overall DG

{

reliability as discussed in Generic Letter 84-15. Therefore, improved TS i

3.8.1 Condition B RA B.3.1 includes the option for making a determination that, upon diesel generator. failure, no common-mode failures exist. This determination precludes unnecessary testing of the remaining diesel generators.

1 5.

Existing TS Survalliance Requirements 4.8.1.1.2.a.5 (one hour load run) i and 4.8.1.1.2.d.6 (hot-restart) do not include any allowances for i

momentary transients outste tha load range during the performance of the 1

surveillance. The improved TS allow momentary transients outside the 1

l load during the corresponding SR 3.8.1.3 and SR 3.8.1.15.

Momentary load transients may occur for various reasons during loading and unloading of i

d the DG. However, these transients are quickly restored to within limits j

j and do not reflect an inability of the system to fulfill its function.

{

Therefore, these transients should not be considered as a failure of the j-surveillance.

a 6.

Existing TS 3.8.1.1 Action d. requires that with an inoperable DG the j

l required feature (s) supplied by the remaining operable DG be verified 4

operable.

If a required feature is determined to be inoperable, the reactor is required to proceed to cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Improved TS 3.8.1 RA B.2 allows redundant feature (s) to be declared inoperable without requiring the plant to be taken to cold shutdown if appropriate actions can be taken. With an inoperable DG and a required redundant feature (s) which is concurrently inoperable and powered by an j

operable DG, certain events that are required by GDC 17 to be capable of being mitigated, may not be able to be mitigated. However, certain i

j combinations of inoperable components may allow for satisfactory compensatory actions. The improved TS allow features to be declared i

inoperable without compelling a plant shutdown, to provide a reasonable j

period of time to take appropriate actions.

7.

Existing TS 3.8.1.1 Actions require that with one DG inoperable the j

redundant feature (s) supplied by thc other DG(s) be verified operable within two hours. The improved TS LCO 3.8.1 RA B.2 extends the time allowed to verify the operability of the redundant feature (s) to four l

hours. The increase will allow the operator sufficient time to evaluate

{

system inoperabilities and take appropriate actions, and minimizes the i

risk due to subjecting the unit to transients associated with shutdown.

This proposed time also considers the capacity and capability of the remaining AC sources and the low probability of a design-basis accident occurring during this period.

8.

The improved TS eliminates the existing TS 4.8.1.1.2 requirement for staggered testing of DGs and supporting features. The intent of the requirement for staggered testing was to increase reliability of the component / system being tested. A number of studies have been performed d

)

i b

e I

=

7.

7..

- -.., _ _ =,

,-,,.g p

9,--

(.

i.

~

130 1

which have demonstrated that staggered testing has negligible impact:on 4

components' reliability. These analytical and subjective walyses have determined that staggered testing is not as significant as -initially j;

thought.

i 9.

Existing TS in Section 3.8.1 require DG starts-without previous engine lubrication, which wears the engine unnecessarily, thereby reducing overall DG reliability. Lubrication does not. contribute to enhanced l

4 engine start performance that could mask the engine's ability to start in

~

accident conditions without a prior lubrication. Therefore,- the licensee proposes a lubrication allowance for each SR requiring a DG start..

1

10. The existing TS 4.8.1.1.2 time requirement for both manually synchronizing the DG to the offsite source and loading of the DG is eliminated in the improved TS. The existing TS requirement does not i

verify any DG capability, only the ability of the operator to synchronize i

power sources and loads. DG loading should.be done in accordance with i

the manufacturer's recommendations to minimize wear on the engine.

1 Additionally, placing a time limitation on the operator to accomplish j

this loading results in an increased potential for error and subsequent j

unavailability of the DG. The testing of DG automatic start and loading capability required by other improved TS SRs is adequate to confirm the i

DG's capability without the manual time requirement.

3

11. The existing TS 4.8.1.1.2 requirement to check and remove accumulated r

water from DG day tanks after each operation of the DG of greater than or l

equal to one hour is eliminated in the improved TS. Checking for and removing accumulated water from the DG day tanks is required every 31 i

days in improved TS SR 3.8.1.5.

Water accumulation within the fuel oil tanks is a time dependent (condensation) process and is not affected by i

the transfer of fuel oil during DG operation.

i

12. The existing TS 4.8.1.1.2 requirements for both auto-starting and loading j.

the DG during the DG hot restart surveillance are eliminated in improved i

TS SRs. Other improved TS SRs verify that the DGs will start and connect l

required safety loads on automatic signals, and it is unnecessary to reaffirm this capability during the hot restart test. The specific

~

requirement for the DG to be automatically loaded with emergency loads.is excessive, since the DG has demonstrated its ability to power loads while hot (i.e., the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> run). Additionally, the automatic loading is an unnecessary repetition of other.SRs which confirm the DG's ' ability to accept loss-of-offsite-power sequenced loads. This revision allows greater flexibility in scheduling DG testing and minimizes unnecessary j

loading transients, while not compromising any necessary demonstration of DG capability.

13.

Existing TS Table 4.8.1.1.2-1, " Diesel Generator Test Schedule," is revised in improved TS Table 3.8.1-1, " Diesel Generator Test Schedule."

4 The improved TS change focuses only on the last 25 tests and eliminates acceptance critoria based on the last 100 tests. This change will yield a more representative reliability of the DG unit at the time of the most j

recent test. Furthermore, the criteria for accelerated testing based on i

... ~..

u...

.m u.--~

2

..u-2 a

w ei.

4 p

i i

1 131 i

the most recent tests is increased from 2-in-20 test failures, to 4-in-25 j

test failures. This increase is consistent with past efforts to reduce testing of the emergency DGs and is considered to provide sufficient data j

for determining whether DG reliability has significantly degraded.

I

14. Existing TS 4.8.1.1.2 requires the DG support system surveillance frequency of performance to be tied to the accelerated test schedule of i

the DG.

Improved TS 3.8.1 includes specific test frequencies for DG i

support equipment and systems, without connection to any DG. accelerated i

test schedule. Verification of DG fuel level, DG fuel transfer pump j

operation, and starting air pressure can be performed on a normal

]

schedule without compromising the DG reliability determinations.

1 15.

In improved TS 3.8.1, the licensee proposes using the phrase " actual or simulated" in reference to the test initiation signals instead of. the j

existing TS 3.8.1 surveillance requirement for verifying that each j

subsystem actuates on a " simulated" or " test" signal.

This proposal clarifies that. satisfactory automatic system initiations for other than i

surveillance purposes can be used to fulfill the surveillance requirements. DG operability is adequately demonstrated in either case i

i since the DG itself can not discriminate between " actual" or " simulated" initiation signals.

l

16. With no offsite AC electrical power available to required features, i

existing TS 3.8.1.2 Actions require the suspension of all Core Alterations, irradiated fuel handling, and OPDRVs.

In lieu'of the existing TS action (s) the improved TS 3.8.2 RA A.1 allows the required feature (s) with no offsite AC power available to be declared inoperable and corresponding RA(s) taken. Since.the improved TS LCO offsite circuit i

operability requirements require power to be supplied to all necessary distribution systems, if one or more required loads, motor control centers, buses, etc. are not capable of being powered via the offsite circuit, the offsite circuit is considered inoperable.

In this event it may not be necessary to suspend all Core Alterations, irradiated fuel l

handling, and OPDRVs. The less restrictive action of declaring affected equipment inoperable (and the associated actions taken) will provide j

. sufficient conservative actions for this condition.

i 17.

Existing TS 3.8.1.2 requires the HPCS DG to be operable when HPCS is 4

required to be operable during shutdown.

Improved TS LCO 3.8.2.c i

provides an option for either the HPCS DG or a second offsite circuit 3

(other than the required offsite circuit-LCO 3.8.2.a) to be capable of j

supplying the HPCS during shutdown. The proposed change will provide an additional flexibility when HPCS is required to be operable.

Since offsite power reliability has proven to be at least as reliable as the typical DG, the use of a second circuit in lieu of the HPCS DG provides

]

' equal or greater power reliability for the HPCS.

18.

Existing TS 3.8.1.2 does not exempt the performance of surveillance 4

requirements requiring the sole required DG to be paralleled to the single required offsite source during shutdown.

Improved TS SR 3.8.2.1-eliminates the performance of DG surveillance requirements that need the 4

1 t

1p

4. -

i 132 I

i one required DG to be paralleled with the one offsite AC source during j'.

shutdown. The paralleling of the only two required.AC power sources presents a risk of a single fault resulting in a station-blackout. The staff has previously recognized this and provided ' surveillance exceptions

}

to avoid this condition.

In an effort to address this concern and to avoid potential conflicting TS requirements, the improved TS allows SRs specified in 3.8.2.1, which would require the DG to be connected to the j

offsite source, and also to be capable of reverting from test back to.

i standby, to be excepted from performance requirements. The exception does not apply to the requirement for the DG to be capable of performing the particular function - just to the requirement to demonstrate such' capability while that source of power is being relied upon to support meeting the LCO.

19. Existing TS 4.8.2.1.c.2 requires verification that the battery connections are " clean, tight, free of visible corrosion." The improved s

TS 3.8.4.5 replaces this with the requirement to remove visible j

corrosion. Removing visible corrosion is' sufficient to ensure the 1

connections are clean. Requiring the' connections to be tight results in 3

a requirement to torque the connecting bolts. The application of a i

torque to confirm tightness results in unnecessary stress being applied 2

to the bolted connection.

If the connection satisfies the resistance i

requirements of improved TS SR 3.8.4.5 (performed at the same frequency 1

as SR 3.8.4.4), it is deemed sufficiently tight for contact.

i i

20.

Existing TS 4.8.2.1.f requires a demonstration of battery life expectancy through an annual performance discharge test of the battery capacity for i

any battery that shows degradation or reaches 85 percent of it: service j

life. A battery can show degradation well before the end of expected i

life, and still be within the required capacity to meet operability-l requirements.

In this event, a frequency of less than 12 months is justified. The 24-month frequency is allowed in improved TS 3.8.4.8 when a battery is at 85 percent of its expected life with a capacity of at l

1 east 100 percent'of the capacity rating by the manufacturer.

l i

21. Existing TS Table 4.8.2.1-1 does not include an allowance for a temporary l

increase in the battery cell electrolyte level when the battery is on j

equalizing charge.

Improved TS Table 3.8.6-1, Note (a), allcws the j

battery cell electrolyte level to temporarily increase above the specified maximum level, if the battery is on equalizing charge.

The e

level excursion is allowed since the equalizing charge can temporarily j

raise tho electrolyte level. The level should return to normal when the battery cells are fully charged. This allowance is' acceptable based on j

guidance from Appendix A to IEEE-450, "IEEE Recommended Practice for Maintenance, Testing and Replacement of Large Lead Storage Batteries for Generating Stations and Substations."

l 22.

Improved TS Table 3.8.6-1 allows charging current limits to be used in place'of existing TS Table 4.8.2.1-1 specific gravity for all categories of battery cell parameters. The existing allowance does not apply to Category B limits. The licensee proposes charging current in lieu of j

specific gravity because the charging current is a satisfactory i

fe -

i i

133 indication of battery cell condition for a limited time after a battery.

is recharged when specific gravity readings may be unreliable.

Limiting i

the time that the charging current may be used in lieu of specific j

gravity will ensure the use of the charging current is sufficiently

]

conservative.

23. An exception to the existing TS Table 4.8.2.1-1, footnote (a),

requirement to correct the specific gravity for electrolytic level is included in improved TS Table 3.8.6-1, Note (b). The exception is allowed by the improved TS, whenever the battery charging current is less than 2 amps when the battery is on float charge. The low charging a

j current provides.an indication that the overall battery condition is good 4

and level correction is not needed.

I

24. The licensee proposes to' include in improved TS Table 3.8.6-1 a general i

statement that charging current may be applied after any recharge. The i

general statement would replace the statement in existing TS Table 4.8.2.1-1 that charging current may be applied after a battery service or i

a performance discharge test.

If the charging current criteria are j

acceptable to monitor the battery status following a deep discharge, that i

same criteria should be acceptable for lesser discharges. This change simplifies the process for determining the' depth of discharge and adjusting the duration for charging current allowance based on that discharge.

1 25.

Existing TS Table 4.8.2.1-1 allows six days to restore Category A and B 1

l limits when batteries are outside Category A limits, and seven days to restore Category B limits when batteries are outside Category B limits.

{

Improved TS 3.8.6 RA A.3 has increased the time allowed for the i

j restoration of battery cell parameters to 31 days. The improved TS l

j increased allowance will not affect the ability of a battery to perform j

its function, since the batteries have the capability to perform design requirements, if the cell parameters are within the Category C limits.

1

(

l 26.

Existing TS 4.8.2.2 requires all battery surveillance requirements to be i

performed on a set frequency and does not consider the effects'of the

]

tests during plant shutdown.

Improved TS SR 3.8.5.1 eliminates the need

]

to perform surveillance requirements which would require an operable i

battery to be rendered incapable of performing its functions during

]

shutdown.

The allowance does not take exception to the requirement that i

. the battery must be capable of performing the particular function - just l

to the requirement to demonstrate that capability while that source of power is being relied on to meet the LCO DC source requirements. The a

i performance of surveillance requirements necessitating that the required I

battery be inoperable presents a significant risk if an event were to i

occur during the test. The staff has previously provided surveillance exceptions to avoid a similar condition for AC sources, but the exceptions have not been applied to DC sources. The change would make the exceptions consistent between the AC and DC sources.

4 27.

Existing TS 3.8.3.2, " Electrical Power Systems, Distribution - Shutdown,"

requires the suspension of all Core Alterations, irradiated fuel j

i I

1 i

a

i o

i 134 handling, and OPDRVs in the event that the required AC or DC distribution I

systems are inoperable. Since the improved TS 3.8.10 distribution system operability requirements require the electrical power distribution subsystem to supply power to all necessary loads, if one or more required loads are not being supplied the required power, the distribution.

subsystem is inoperable.

Improved TS 3.8.8 RA A.1 allows associated supported required feature (s) to be declared inoperable in lieu of other ras such as the suspension of all Core Alterations, irradiated fuel handling, and OPDRVs.

By allowing the option to declare required features associated with an inoperable' distribution subsystem inoperable, appropriate restrictions are implemented in accordance with the affected distribution subsystem LC0's Required Actions.

In this event it may not j

be necessary to suspend all Core Alterations, irradiated fuel handling, and OPDRVs. Conservative actions can be assured during shutdown, if all required features without electrical: power can be declared inoperable and the associated actions taken for the features in lieu of suspension of.

operations.

28. Existirg TS 3.8.2.2 Actions require the.suspens' ion of all Core Alterations, irradiated fuel handling, and OPDRVs in the event that a required DC source is inoperable during shutdown. Since improved TS 3.8.5 now requires DC sources to be supplying all required distribution systems, the improved TS includes an alternate Action when loads are without power.

Improved TS 3.8.5 RA A.1-allows the affected features to be declared inoperable in lieu of completing all other ras,' including the suspension of all Core Alterations, irradiated fuel handling, ar.d OPDRVs.

By allowing the option to declare required features inoperable with associated DC power source (s) inoperable, appropriate restrictions are implemented in accordance with the affected system.LCO's Actions.

In this event it may not be necessary to suspend all Core Alterations, irradiated fuel handling, and OPDRVs. Conservative actions can be assured if all required equipment without the aecessary DC power is-declared inoperable and the associated actions taken.

29. Existing TS 4.8.2.1.b.3 requires that whenever the battery experiences a discharge or overcharge, verification must be made that the average electrolyte temperature of the pilot cells and representative cells of connected cells is above 65'F.

Improved TS SR 3.8.6.3 requires verification of average electrolyte temperature of representative cells once every 92 days and removes the specific requirement to verify the cell temperature after a battery discharge or overcharge. However, improved SR 3.8.6.2 continues to require that battery cells meet Category B limits after specified battery overcharge. Therefore, the temperature i

verification requirement in the existing TS can be deleted without adverse affects on plant safety.

30. Existing TS 4.8.2.1.b.2 requires tMt within seven days after a battery discharge or overcharge, a visual maination be performed to determine that "there is no visible corrosin et either terminals or connections."

Battery terminal corrosion is a 16 term process and a battery discharge or overcharge does not induce corrosion.

Improved SR 3.8.4.2 deletes the link to battery discharge or overcharge and only requires -verification of-

_ _ _. ~. _. _, _ _ _. _

b l

135 no visible corrosion at battery terminals and connectors or a resistance l

check once every 92 days. Both the existing TS and improved TS require l

4 all battery cell connections to be cleaned and coated with an anti-corrosion compound every 18 months. The 18 month SR is sufficient to determine the condition of the terminals.

i i

j 31.

Existing TS 3.8.3.1 requires the licensee to restore an inoperable AC j

load-shedding and load-sequencing panel (automatic load sequencer) to operable status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Improved TS 3.8.1-RA F.1 would allow 24 j

hours to restore an inoperable automatic load sequencer to operable.

l status. The improved TS change is justified because although the safety 4

J loads for one division could not be automatically loaded, the loads can '

4 be manually loaded when required.

i The above less restrictive requirements are acceptable because they do not

{

present a significant safety question for the operation of the plant. The remaining TS requirements are consistent with current licensing practices, i

operating experience, and plant accident and transient analyses. They give

)

reasonable assurance that the public health and safety will be protected.

}

)

l 3.9 Refueling Operations l

The licensee has adopted the NUREG-1434 specifications for Section 3.9 except i

for improved technical specification (TS) 3.9.8 " Residual Heat Removal (RHR) -

High Water Level" and 3.9.9 " Residual Heat Removal - Low Water Level." These j

specifications have been modified slightly to account for plant specific decay heat removal systems and containment design characteristics.

i j

A.

Significant Administrative Changes.

In accordance with the guidance in the Final Policy Statement, the licensee has proposed administrative changes to the existing TS to bring them into conformance with the improved TS. These changes are as follows:

3 I

1.

Existing TS 3.9.1, " Reactor Mode Switch," has been divided into two 1

separate limiting conditions for operation (LCO) in the improved TS.

i Improved TS 3.9.1, " Refueling Equipment Interlocks,". specifies the l

conditions, required actions and surveillances associated with the j

refueling equipment interlocks that were specified in existing TS 3.9.1.

Improved TS 3.9.2, " Refuel Position One-Rod-Out Interlock," specifies -

i conditions, required actions and surveillances. associated with the one-l rod-out interlock. This division of requirements is an administrative i

change with the requirements effectively retained within the improved TS and is therefore acceptable.

i 2.

During Mode 5 with an accumulator associated with a withdrawn control rod 4

inoperable, existing TS 3.1.3.3 Action b.1 and improved TS 3.9.5 Required -

4 Action A.1 require that the control rod be inserted. Once the control i

rod is fully inserted, the accumulator is no longer required to be operable per existing TS Footnote

• and improved TS 3.9.5.

In addition, entry conditions for the required actions are no longer applicable, so 4

i

1 P'

4 i

+

i i

136 j

i that no additional actions are required. This is consistent with both the existing TS and improved TS LCO 3.0.2.

Thus, the action to disarm the associated directional control valves is unnecessary, and has been

)

I deleted. The staff finds this deletion acceptable.

3.

Existing TS 3.9.11.1 Action a. and TS 3.9.11.2 Action a. requires that if l

an alternate method of decay heat removal cannot be demonstrated for the 3

inoperable RHR shutdown cooling subsystem,' all' operations' involving an increase in the reactor decay heat load are to be suspended.

Improved TS 3.9.8 requires only that loading of irradiated fuel assemblies into the i

reactor pressure vessel be suspended since this is the only practical.

i 4

method of increasing the reactor decay heat load given the applicability j

and conditions of the existing and improved TS. Additionally, improved TS 3.9.9 does not include this action since loading of irradiated fuel j

assemblies is not allowed in the applicable conditions by improved TS l

l 3.9.6.

Since these requirements result in the same response as the i

current requirements, the change is an administrative change and is j

acceptable.

i l

4.

The existing TS 3.9.11.1 Action a. requirement to establish containment i

integrity when a loss of RHR shutdown cooling occurs during fuel handling l

l could allow 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during which integrity could be violated even if-capable of being established. The appropriate required actions are j

provided in improved TS 3.9.8.

These actions allow one hour to verify alternate methods of decay heat removal, otherwise, immediate i

j requirements are imposed to establish and maintain the containment -

l boundary.

Improved TS 3.9.8 required Action B is modified to account for the plant j

specific containment design and the required boundary for this condition as defined by the current licensing basis. To accommodate these

,i variations, plant specific wording is used in the LCO RA and Bases. Just L

as during power operations, some penetrations may be administrative 1y i

open in mode 5 to support RHR operability requirements during high water level conditions. Since not all penetrations are assumed to be isolated i

or capable of being isolated because the penetration may utilize j

isolation devices other than valves; only those that are required (i.e.,

[

assumed to perform a containment isolation function) are necessary to j

satisfy the containment integrity requirements. These same changes have i

been made to improved TS 3.9.9.

These changes are plant specific and conform to the current licensing basis and they are acceptable.

5.

Existing TS 3.9.11.1 and 3.9.11.2.specify that either the RHR system or the alternate decay heat removal system (ADRS) shall be operable and in operation when reactor cavity water level limits are not met.

Improved a -

TS 3.9.8 specifies that one RHR system is required to be operable and in operation for high water level and LCO 3.9.9 specifies that two RHR systems are required to be operable with one in operation for low water i

level. The licensee has modified improved TS 3.9.8 and 3.9.9 to take i

advantage of their ADHRS design and current licensing basis. The staff i

has reviewed the changes to improved TS 3.9.8 and 3.9.9 and finds that l

i 4

i i

- ~.. -, _ _ - -. -

i i

137 4

j the changes are in accordance with the current licensing basis and are acceptable.

i l

6.

Existing TS 3.9.11 action b. specifies requirements for no RHR shutdown cooling system operable.

In the existing TS 3.9.11.1 and 3.9.11.2, either of the installed decay heat removal systems, i.e., the RHR i

shutdown cooling subsystem or the ADHRS, satisfies the requirements of.

the LCO. The current Action b., however, is inconsistent with the LC0 in 3

that it does not recognize that the ADHRS could be the decay heat removal system in operation in accordance with action b.

This statement has been' revised in improved TS 3.9.8 and 3.9.9 for Condition C to be consistent with the LCOs and is acceptable.

e l

B.

Relocated Requirements In accordance with the guidance in NUREG-1434, the licensee has proposed to relocate all or portions of the following existing'TS within the improved TS:

Existina TS Title 3/4.9.1 Reactor Mode Switch - Surveillance Requirements 3/4.9.2 Instrumentation i

3/4.9.9 Water Level-Spent Fuel Storage'and Upper Containment Fuel Pools 3/4.9.10.1 Single Control Rod Removal 4

i 3/4.9.10.2 Multiple Control Rod Removal l

The more significant changes resulting from relocated items are as follows:

i 1.

The note associated with existing TS 4.9.1.2 and 4.9.1.3 specifies the l

position of the reactor mode switch and the surveillances required for the control rods when testing the reactor mode switch interlocks. The j

improved TS address these items within specification 3.10.2 " Reactor Mode i

Switch Interlock Testing." Thus, the interlock conditions for testing in i

the existing TS are effectively retained within the improved TS.

I 2.

Existing TS 3/4.9.2, 3.9.10.1.b, 3.9.10.2.b, 4.9.10.1.b and 4.9.10.2.b i

states the operability and surveillance testing requirements for source i

range monitors in Mode 5.

The improved TS address these requirements within Specification 3.3.1.2 " Source Range Monitor (SRM)

Instrumentation." Thus, the SRM operability requirements in'the existing TS are effectively retained within the improved TS.

1 3

3.

Existing TS 3/4.9.9 specifies the minimum water level in the spent fuel-storage and upper containment fuel pools and the surveillance frequency i

for monitoring water level. The improved TS address these items within Specification 3.7.7 " Fuel Pool Water Level." Thus, the fuel pool water 1evel requirements in the existing TS are effectively retained within the improved TS.

1 4.

Existing TS 3/4.9.10.1 specifies the conditions, actions, and surveillances associated with the removal of single control rods when in i

Modes 4 and 5.

Existing TS 3/4.9.10.2 specifies the same items for 3

_j

4 f

138 L

multiple control rod removal when in Mode 5.

The improved TS address the majority of these requirements within specifications 3.10.4 " Single

}

Control Rod Withdrawal - Cold Shutdown," 3.10.5 " Single Control Rod Drive (CRD) Removal - Refueling," 3.3.1.2, "SRM Instrumentation," and 3.10.6:

i

" Multiple Control Rod Withdrawal - Refueling." The. remaining items have been relocated to plant controlled procedures. Thus the control. rod 4

requirements for Modes 4 and 5 in the existing TS are effectively retained within the improved TS.

j 5.

Existing TS 3.9.10.2.c and 4.9.10.2.1.c specify the Shutdown Margin (SDM) requirements in Mode 5.-

The improved TS address these requirements 1

within Specification 3.1.1 " Shutdown Margin." Thus the SOM requirements in the existing TS for Mode 5 are effectively retained within the

)

improved TS.

i j

6.

Existing TS 3.1.3.3 Mode 5 requirements for multiple control. rod l

withdrawn while fuel remains in its cell is only accomplished in

]

accordance with improved TS 3.10.8, "SDM Test - Refueling." In the event 4

multiple withdrawn control rods have inoperable accumulators and no CRD i

pump operating (rewritten as " charging water header pressure C 1520 J

psig"), the actions of improved TS 3.10.8 will direct the immediate 2

scram. Actions for multiple withdrawn control rods with' inoperable accumulators and CRD pump operating, will.be addressed by the improved TS i

l 3.9.5.

The result is operation and actions consistent with existing TS j

requirements.

t l

The above changes are considered administrative changes in the location of the 1

requirements within the TS and are therefore acceptable.

a In accordance with the guidance in the Final Policy Statement, the licensee has proposed to relocate or reorganize all or portions of the following i

existing TS to other licensee-controlled documents l

Existina TS Title i

j 3/4.9.1 Reactor Mode Switch

{-

3/4.9.4 Decay Time 3/4.9.5 Communications 3/4.9.6.1 Refueling Platform 3/4.9.6.2 Auxiliary Platform 3/4.9.7 Crane Travel 1

3/4.9.8 Water Level Reactor Vessel j

3/4.9.11.1 Residual Heat Removal and Coolant' Circulation -

g High Water Level 3/4.9.11.2 Residual Heat Removal and Coolant Circulation -

t Low Water Level j

3/4.9.12 Horizontal Fuel Transfer System j

The more significant changes resulting from relocated items are as follows:

l 1.

Existing TS 3/4.9.1, 3/4.9.10.1.a, 3/4.9.10.2.a and 4.9.10.2.1.a specify j

that the reactor mode switch shall be operable and locked in the shutdown i

or refuel position for Modes 4 or 5 operation.

Reactor mode switch i

I

.sm..

-.-a a--

-w.

u a

a,

-.~.,

kl=

i j.

139 I

operability is included as part of the operability of the associated t

interlocks required by improved TS 3.9.1 and 3.9.2.

Movement of the reactor mode switch from the shutdown position is adequately controlled 4

i by the improved TS Modes definition Table 1.1-1.

Thus, the requirement to " lock" the mode switch is redundant and unnecessary, and the locking i

requirement can be adequately controlled outside of.TS. The provision to lock the reactor mode switch in Modes 4 or 5 will be relocated to the-TS

{

Bases.

I j

2.

Although Criterion 2 of the Final Policy Statement would require existing TS 3/4.9.4 " Decay Time" to be retained in the improved TS, the 4

i requirement for a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay time following suberiticality before.

commencing movement of irradiated fuel In the reactor vessel will always be met for a refueling outage. The operations required prior to moving irradiated fuel in the reactor vessel (e.g., containment entry, removal of drywell head, removal of vessel head, removal of vessel internals)be j

j require in excess of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to complete before irradiated fuel can moved. Therefore, the requirement is unnecessary and has been relocated

[

from the specifications to the UFSAR.

t i

3.

Existing TS 3/4.9.5 requires that communication between the control room-and refueling floor personnel be maintained to ensure that refueling l

personnel can be promptly informed of significant changes in the plant t

j status or core _ reactivity conditions during refueling. The refueling t

i system design accident or transient response does not take credit for communications and is only designed to ensure safe refueling operations.

{

Therefore, the requirements have been relocated to the UFSAR and will be j

controlled in accordance with 10 CFR 50.59.

4.

Refueling platform operability specified in existing TS 3/4.9.6.1 and i

3/4.9.6.2 ensures that appropriate controls are in place for handling of radioactive components and core internals. Although interlocks are i

designed to prevent damage to these components, the interlocks are not l

relied upon to prevent or mitigate the consequences of a design basis j

accident. Therefore, the requirements have been relocated to the UFSAR or TS Bases and~will be controlled in accordance with 10 CFR 50.59.

?

5.

The crane travel limits specified in existing TS 3/4.9.7 and the action i

statements of existing TS 3.9.9 are provided by physical design and j

administrative controls. Although these specifications support the l

maximum refueling accident assumptions, the fuel handling crane travel limits are physical design limits and not process variables which are monitored and controlled by the operator. Therefore, the requirements i

have been relocated to the UFSAR and TS Bases and'will be controlled in j

accordance with 10 CFR 50.59.

i i

6.

The applicability of existing TS 3.9.9 is being changed in the improved =

TS from "when irradiated fuel assemblies are in the fuel storage pool" to "During movement of' irradiated fuel' assemblies in the associated fuel i

storage pools.' The bounding design basis fuel handling accident over the spent fuel storage pool assumes an irradiated fuel assembly is j

dropped onto an array of irradiated fuel assemblies seated within the i

l'

_=-

1

=

f i

4 l

140 t

pool racks. Because 60% of the damaged fuel rods are in the dropped-assembly, the consequences for any other event would be significantly i

1 j.

reduced. As indicated above, administrative controls specified in existing TS 3/4.9.7 and 3.9.9' Actions are relocated out of the TS. The-j additional limits expressed in existing TS 3.9.9 for water level during j

periods other than fuel handling operations are relocated to the UFSAR.

4

{

7.

Existing TS 4.9.11.1 and 4.9.11.2 specify the method (s) of verifying the operation of the decay heat removal systems during Mode 5 operation.

j Although maintaining decay heat removal capability was identified as an i

important contributor to risk reduction and is still retained in the i

improved TS, specifying the methods for determining that capability

]

depends on plant conditions and operator actions and can be adequately j

controlled by the improved TS Bases.

8.

The fuel transfer system specified in existing TS 3/4.9.12 provides a method of moving fuel between the fuel building and the primary j

containment without high radiation exposure levels for personnel.

The system is not relied upon to prevent or mitigate the consequences of a design' basis accident but is provided to maintain personnel exposure i

below the limits of 10 CFR Part 20 for normal operation. 'Therefore, the i

requirements have been relocated to the UFSAR and will be controlled in I

accordance with 10 CFR 50.59.

j 9.

Existing TS 3.9.2 and 4.9.2.a.2 specify verifying that the SRM is inserted to the normal operating level. The methods for determining the location of the SRM within the reactor' depend on plant design features and system operation. The design features and system operation which j

dictate the methods identified in the existing TS are described in the i

UFSAR. Since the procedural details of how a specific surveillance is performed are not to be located in the improved TS under NUREG-1434 and i

are not relied upon to prevent or mitigate the consequences of a design basis accident, the methods can be relocated to the Bases and removed l

from the TS.

i i

10.

Existing TS 3.9.10.1.d specifies the conditions and surveillances for the i

other control rods associated with the single control rod being withdrawn. These specifications also.specify the methods in detail for disarming the control rods. The methods for disarming the control rods.

j.

depend on plant conditions and operator actions.

In addition, the i.

procedural details of how a specific surveillance is performed are not located in the improved TS under NUREG-1434 and are not relied upon to prevent or mitigate the consequences of a design basis accident, the methods can be relocated to the Bases and removed from the TS.

I

11. Auxiliary platform operability specified in existing TS 3.9.6.2 and 3.9.6.3 ensures that appropriate controls are in place for handling of radioactive components and core internals. Although interlocks are i

designed to prevent damage to these components, the interlocks are not relied upon to mitigate the consequences of a design basis accident (fuel i

handling accident). Therefore, the requirements specified in existing TS

}'

e i

j j

141 3.9.6.2 and 3.9.6.3 have been relocated to the UFSAR and will be controlled in accordance with 10 CFR 50.59.

12.

Existing TS 3.9.11.1 and 3.9.11.2 require ~the one operable RHR shutdown cooling subsystem to have an associated operable diesel generator. This requirement was specifically added to provide for additional capability 1

for mitigation of a loss of decay heat removal event during use of the

]

alternate decay heat removal system (ADHRS). However, the existing TS allow for unlimited continued operation following the loss of the RHR shutdown cooling system using any alternate decay heat removal method.

l Therefore, the change to allow ADHRS to be used to satisfy the LCO did not reduce the capability to mitigate the loss of decay heat removal due-to a loss of offsite power for the allowed conditions controlling unlimited continued operation.

i t

Generally, outage activities are planned so that equipment required to be

}

operable during a shutdown is powered by the one. required operable diesel i

generator, but this is not a TS requirement. The improved TS allow j

required equipment to be powered by either Division 1 or 2, or split between the divisions without regard for which diesel generator is j

operable. This-is based on significantly reduced energy contained in the i

reactor coolant pressure boundary, very low reactor coolant operating '

temperatures and pressures, and minimal corresponding stresses which result in probabilities of occurrence that are.significantly reduced'or j

eliminated, and with minimal consequences should they occur.

4 Although the improved TS do not require that the operable RHR shutdown i

cooling system be powered by an operable associated diesel generator,' the licensee has committed to maintain this equipment arrangement during i

i shutdown conditions, i.e., Modes 4 and 5.

However, this commitment can i

be administrative 1y maintained as is currently being done at other plants. Therefore, the control of this commitment under 10 CFR 50.59 is l

acceptable and it can be relocated to TS Bases.

i

13. Control rod movement testing (control rod " notching" and coupling checks), requires control rod position indication to be operable.

If position indication is inoperable, then testing required by improved TS 3.9.4 can not be satisfied and appropriate actions are required for j

inoperable control rods, due to failed SRs.

Existing TS SRs 4.1.3.5.b and c represent details found within existing procedures for the l

performance of TS surveillances. Therefore, relocating the details of current surveillance 4.1.3.5.b to the TS Bases and control of this commitment under 10 CFR 50.59 is acceptable.

j 14.

Existing TS SR 4.1.3.3 requirements for measuring accumulator check valve 4

performance by recording the time the accumulator pressure remains above j

the alarm setpoint with no control rod drive pump operating is not assumed in accident or transient analyses. With no operating control rod j

drive pump, the reactor must be scrammed within a short time if a pump is j

not restored to operation (no more than 20 minutes after two or more t

1 e

,,,m.,

c._,.

._s.

._,....,,,...w

,___._.m..

,,.. ~,,

....o

,,,_.,mmma,,._._m..

N a

I j

142 i

accumulators have low pressure). Therefore existing TS SR 4.1.3.3'for

].

control rod operability in mode 5 is deleted from TS and controlled by-10 CFR 50.59.

4 In addition the scram accumulator leak detectors, pressure detectors,.and l

associated alarms do not necessarily relate directly to accumulator I

J operability.

In general the BWR Standard Technical Specifications, NUREG-1434, does not specify indication-only or test equipment to be operable or to support operability of a system or component. Control of j

this equipment can be adequately addressed by plant operational procedures. Therefore, the control of this commitment under 10 CFR 50.59 is acceptable and it can be relocated to UFSAR or TS Bases.

F

{

The above relocated requirements relating to refueling operations are not required to be in the TS under 10 CFR 50.36, and are not required to obviate l

the possibility of an abnormal situation or event giving rise to an immediate i-threat to the public health and safety. Further, they do not fall within any j

of the four criteria set forth in the Commission's Final Policy Statement, j

discussed in the Introduction above.

In addition, the staff finds that sufficient regulatory controls exist under 10 CFR 50.59. Accordingly, the i.

staff has concluded that these requirements may be relocated from the TS to the licensee's TS Bases or the UFSAR, as applicable.

(

i l

C.

More Restrictive Requirements i

By electing to implement the NUREG-1434 Section 3.9 Specifications, the j

licensee has adopted a number of more restrictive conditions than are required by the existing TS. The more significant conditions are the following:

1 1.

The existing TS 3/4.1.3.5 actions for inoperable control rod position j

indication in Mode 5 only require insertion of the control rod. The improved TS 3.9.4 actions require, in addition to inserting the control i

rod, that fuel movement and control rod withdrawal be suspended (Required Action A.1.1 and A.I.2), or that the inserted control rod be disarmed (Required Action A.2.2).

This prevents additional core reactivity changes while actions are being taken to insert the control rod with the inoperable position channel.

The alternative actions require immediate initiation of insertion of the control rod associated with the inoperable i

position channel and disarming of the associated fully inserted contrc'1 rod drive. These actions ensure that the control rod associated with the i

inoperable position channel cannot be withdrawn, thus precluding two i

control rods from being inadvertently withdrawn due to control rod position channel failure. Finally, the proposed completion time has been added to specify that the Required Action be completed "immediately," in l

contrast to the existing TS action which does not clearly specify a time period to start or complete the action.

2.

Existing TS 3/4.1.3.3 " Control Rod Scram Accumulators" is rewritten in improved TS 3.9.5 " Control Rod Operability to require control rod -

i Operab'ility instead of only CRD accumulator Operability.

Further,-the improved TS adds an additional requirement that each withdrawn control

{

rod must be capable of insertion (improved TS SR 3.9.5.1).

This added

i i

]

143 requirement, in conjunction with the existing accumulator requirement,

- i provides the appropriate assurance that the withdrawn control rod can be manually inserted as well as inserted on a scram signal if ' required.

l 3.

The requirement of existing TS 3.9.11.1 Action a. requires that with less i

than the required RHR systems operable, within one hour and at least one j

per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter, the operability of at least one alternate method j-of decay heat removal shall be demonstrated.

In the event this cannot be accomplished, the-action requires establishing containment integrity t

within_4 hours. This would appear to provide a period of time (four hours) during which integrity could be violated even if capable of being i

maintained. Additionally, if the plant status is such that integrity is i

not capable of being established within four hours, the existing TS~

action results in non-compliance with the TS and a requirement for a licensee event report. The intent of the action is believed to be more appropriately presented in the Required Actions of improved TS 3.9.8.

These actions impose a significantly more conservative requirement to establish and maintain the containment boundary. No longer would the provision allowing the plant to violate the boundary for up to four hours appear to exist, based on the existing TS wording. The licensee modified improved TS 3.9.8 required Action B to account for the plant specific containment design and the required boundary for this condition as defined by the current licensing basis.

l t

4.

The existing TS 3.9.11.2 action to " initiate action within one hour" to i

establish containment integrity is proposed to be revised to " initiate action immediately." The existing action would appear to provide a period of time during which integrity could be violated even if capable i

of being established. The intent of the action is believed to be more

+

appropriately presented in the improved TS ras. With the improved TS, a significantly more conservative requirement to begin efforts to establish and maintain the secondary containment boundary is imposed.

The staff has reviewed the above more restrictive requirements and concludes that they result in an enhancement to the improved TS. Therefore, the more restrictive requirements are acceptable.

D.

Less Restrictive Requirements The licensee in electing to implement NUREG-1434 Section 3.9 Specifications has proposed a number of less restrictive conditions than are allowed by the existing TS. The more significant conditions are the following:

1.

Existing TS 4.9.1, 4.9.3, and 4.9.8 require that in addition to the normal periodic surveillance frequency, an additional surveillance be performed prior to use. The normal periodic surveillance frequency for the component or system provides adequate assurance of operability. As such, the requirement to perform the surveillance requirement "within X hours prior to" use of the component or system has been deleted.

If the surveillance has not been performed within the specified interval, use of -

the component is not allowed since improved TS SR 3.0.4 and existing TS 4.0.4 require that a surveillance be performed pr_ior to entering the r

,,,m__,._.-.--~+.,+..-w.

. * + = * * - * ' " ~ ' ' * * ~ - - " " *

]'

.c; I

i 144 j

i applicable Mode or condition and be current. This ensures the requirements are adequately checked prior to and.during the operations l

specified in the improved TS.

I 2.

Existing TS 3.9.1 Action b. requires the reactor mode. switch to be locked j

in the shutdown position when the one-rod-out interlock is inoperable.

This action is revised in improved.TS 3.9.2 to immediately. suspend control rod withdrawal and initiate action to insert all insertable control rods in core cells containing one or more fuel assemblies. These actions compensate for an inoperable one-rod-out interlock:and provide j

j adequate protection against potential reactivity excursions.

Further,

-l l

moving the mode switch to the shutdown position could cause an-unnecessary pressure transient on the control rod drive system.

i 3.

Existing TS 4.9.1 specifies the surveillance to be performed on the reactor mode switch interlock. 'To properly perform a test of'the one-i rod-out interlock, a control rod must be withdrawn.. However,' existing TS l

4.0.4. and improved TS SR 3.0.4 prohibit entry into the. applicability of an LCO unless its required surveillances are performed. Therefore, an f

allowance is.provided to enter the improved TS LCO 3.9.2 applicability for a short time (one hour) to provide adequate time to perform the improved TS SR 3.9.2.2.

The one hour frequency.is. considered adequate because of the procedural controls on control rod withdrawals.

In j

addition, indications are available in the control room to alert the l

operator of any control rods that may not be fully inserted.

4.

The requirement in existing TS 3.9.3 for all control rods to be inserted j

during Core Alterations is revised in improved TS 3.9.3 to be fully inserted when loading fuel assemblies into the core. This is consistent r

j with the accident analysis. The control rod withdrawal error during refueling analysis assumes all control rods are inserted only during fuel

(

loading, not unloading. A fuel unloading error cannot increase the I

reactivity of the core and cause an inadvertent criticality. Therefore,.

i the applicability of improved TS 3.9.3 has been specifically tied to i

loading fuel assemblies into'the core consistent with accident analysis j

assumptions, and the Required Actions have been revised to reflect placing the plant in a condition in which the LCO does not apply.

5.

The requirement in existing TS 3/4.1.3.5 requires the position indication i

system for Mode 5 to indicate the current position of the control rod.

~

This position indication requirement is effectively omitted in improved l

TS 3.9.4 in that no position indication is required other than " full-in."

l The operability of the control rod " full-in" position indication for each j

control rod is required to support operability of the refueling interlocks (improved TS 3.9.1) and the operability of the one-rod-out interlock (improved TS 3.9.2).

The previous requirement to provide position indication was only provided for operator information and was not needed other than to indicate whether or not the control rods were j

" full-in."

r

).

Improved TS 3.9.4 also adds a specific requirement for one " full-in" j

position indication channel to be operable for each control rod -

j t

145 i

regardless of the actual position of the control rod.

This added

{

restriction details requirements consistent with the intent of requiring s

the refueling interlocks and the one-rod-out interlock to be operable.

l Improved TS 3.9.4 and 3.9.5 for Mode 5 do not require the specific i

position of a withdrawn control rod to be indicated, and therefore, actions are also not required if control rod position is unknown. The

.I j

proposed requirement would only require that a withdrawn control rod not i

i indicate " full-in". Consistent with this discussion, existing TS 4.1.3.5 i

is also modified to require only that the full-in indication channel be i

operable. Since only one control rod can be withdrawn while in Mode 5 (exceptions to this are addressed in improved TS - Section 3.10), and the 1

1 l

position of the control rod is not a consideration'in any accident or j

transient in this condition, the precise position of the control rod is i

only an operational consideration, not a safety concern.. The critical j

safety issue, whether the control rod is' fully inserted or not, is j

addressed by the improved TS 3.9.4.

6.

Existing TS 3/4.9.8 has been reformatted into two specifications (improved TS 3.9.6 and 3.9.7).

Improved TS 3.9.6 provides the requirements for movement of irradiated fuel assemblies within the Reactor Pressure Vessel (RPV) with water level. determined from the top of the RPV flange. This is consistent with existing TS 3/4.9.8.

Improved j

TS 3.9.7 provides the requirements for movement of new fuel assemblies or handling of control rods within the RPV when irradiated fuel assemblies are seated within the RPV. The water-level is determined from the top of j

irradiated fuel assemblies seated within the RPV rather than from the top of the RPV flange as in the existing TS.

The change to water level measurement is based on there being sufficient water necessary to retain iodine fission product activity in the water in the event of a fuel handling accident. The fuel handling accident would occur at the top of 1

the irradiated fuel seated within the RPV since a new fuel assembly or j

control rod would not contain fission products and, therefore, would not create a release of fission products assumed in the safety analysis at a

l the flange.

I i

i 7.

Existing TS 4.1.3.5.d requires a channel check of the control rod j

position indication system on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> frequency when the alternate j

control rod position indicator is operable. The Rod Pattern Control System receives control rod position inputs from two independent channels.

The system continuously performs a self test to confirm i

agreement between the two channels (i.e., channel check). Should either 4

j channel's position input become unknown, should the two channels disagree i

on position, the channel check otherwise fail, or the test circuit fail i

to perform any self test, the system initiates a control rod block and j

alarms the malfunction. A 12-hour channel check when the alternate.

control rod position indicator is operable is therefore superfluous.

Deleting this existing TS surveillance by not including it.in improved TS i.

3.9.4 will not have an impact on the reliability of the system, on the ability of the operator to detect a control rod whose. position is unknown, or on safe operation of the plant and is acceptable.

a'

j 1

j' e

i 146 i

8.

Any time the operability of existing TS 3/4.9 system or component has

[

d been affected by repair, maintenance or replacement of a component, as discussed in the Bases for SR 3.0.1, post maintenance testing is required 4

i to demonstrate operability of the system or component.. Explicit post l

maintenance SRs have therefore been deleted from these specifications.

l The above less restrictive requirements have been reviewed by the staff and l

have been found acceptable because they do not present a significant safety-question in the operation of the plant. The TS requirements that remain are-consistent with current licensing practices, operating experience, and plant i

accident and transient analyses, and provide reasonable assurance that the l

public health and safety will be protected.

l l

3.10 Special Operations The licensee has accepted the NUREG-1434 Specifications for Section 3.10 j.

except for LCO 3.10.9, " Recirculating. Loops - Testing," and LCO 3.10.10

" Training Startups" which is no longer required. A number of surveillances and clarifying notes that specify methods or procedures for verifying j

equipment and system operability and conditions which are not contained in the improved TS, have been relocated to licensee-controlled documents. These 4

j items are contained in the existing TS which have been relocated to Section 3.10, and are discussed in the applicable' discussion above pertaining to the

{

section from which they were relocated in the existing TS.

4 A.

Significant Administrative Changes In accordance with the guidance in the Final Policy Stater. ant, the licensee has proposed administrative changes to the existing TS to bring them into j

conformance with the improved TS. These changes are as follows:

1 1

1.

Existing TS 3.1.3.4 requirements for control rod coupling in Mode 5 have been relocated to improved TS 3.10.8 as specific restrictions for l

Shutdown Margin (SDM) demonstrations in Mode 5 since they are deleted as j

normal Mode 5 requirements. This change also includes the appropriate actions and surveillances. As a result of this change, the time allowed i

. for the action - insert control rod and disarm - is changed from two hours to four hours.

In Modes 1 and 2 the existing action for an uncoupled control rod (existing TS 3.1.3.4, Action a.2) allows two hours before entering the existing TS LCO 3.1.3.1 Action b.1, which then gives an additional hour to insert and disarm the control rod - for a total of i

three hours to insert and disarm. This existing three hour allowance, before requiring that an inoperable (uncoupled) control rod be inserted, is the time found in improved TS 3.1.3 Required Action C.1 for control i

rod insertion. For consistency of presentation, this three hour limitation is also proposed for all other instances of inoperable control j

rods.

These other instances also warrant a minimal time to attempt restoration prior to inserting and disarming. Given that these instances do not represent loss of SDM, the extended time does not represent a i

significant safety concern. Thus, the change is considered an i-administrative change, and is acceptable.

l'

d I

147 2.

Existing TS 4.1.3.4.a addresses the requirement to perform coupling I

checks after performing activities which could have affected coupling integrity. This surveillance must be completed prior to allowing the l

control rod to be considered operable.

Improved TS SR 3.10.6.5 i

editorially rewrites TS 4.1.3.4.a to more clearly define control rod operability. The rewritten requirement results in existing TG 4.1.3.4.c i

becoming a subset of the existing TS 4.1.3.4.b and..c requirements. The

. performance of.the integrity verification prior to control rod operability (improved TS SR 3.10.8.5) bounds the " prior to reactor 4-criticality" of existing TS 4.1.3.4.c.

The elimination of the existing SR' represents no actual change in requirements. Elimination of existing TS 4.1.3.4.c is thus an administrative change and is acceptable.

3 3.

Existing TS 3.9.10.1.c.1 and 3.9.10.1.c.2 work together, referring to an exception to the current normal SDM requirements, which requires-additional margin for immovable control rods.

Improved TS LCO 3.10.4.c.2 and LCO 3.10.5.c do not include the existing c.2 wording, but identify i

that the withdrawn rod is considered to be the " highest worth control rod" which is assumed in the definition of SDM to be fully withdrawn.

Since the rod is only considered once in the SDM calculations, this rod i

is not required to also be considered as a stuck rod in accordance with the definition; therefore, the existing additional wording is i

unnecessary, and its deletion is acceptable.

t i

4.

The requirement in existing TS 3.9.10.1.d which states "The four fuel assemblies surrounding the control rod or control rod drive' mechanism to i

be removed from the core and/or the reactor vessel are removed from the core cell" is not included in improved TS LCO 3.10.4 and LCO 3.10.5.

Improved TS 3.10.4 is applicable in Mode 4.

In this mode the requirement cannot be physically met and therefore its deletion is acceptable.

Improved TS 3.10.5 is applicable in Mode 5.

During Mode 5 if the four i

fuel assemblies in a core cell are removed, the associated single control i

rod can be removed under the provisions of existing TS 3.9.10.2 (improved l

TS 3.10.6). The limitations of existing TS 3.9.10.1 are consistent with the limitation in existing TS 3.9.10.2 for this situation. Since the multiple withdrawal specification is adequate to control this condition, i

this option is not included in improved TS 3.10.5, and its deletion is acceptable, a'

5.

The licensee has elected not to taL advantage of a number of existing TS in the improved TS. The licensee is deleting existing TS 3.10.1 " Primary Containment Integrity /Drywell Integrity," existing TS 3.10.4 i

" Recirculation Loops" (improved TS 3.10.9 " Recirculating Loops-Testing"),

i existing TS 3.10.5 " Training Startups," (improved TS 3.10.10. " Training l

Startups" and the portion of existing TS 3.10.2 " Rod Control Pattern i

System" (improved TS 3.10.7 " Control Rod Testing - Operating") which refers to startup tests.

Since all low power physics and startup tests have been completed, the licensee no longer needs these exceptions. The staff concurs that these specifications are no longer needed since the

{

low power physics and startup tests have been completed, and finds the deletions acceptable.

4 E

p

.y, w

y.

,,m,..

1i i

148 i

B.

Relocated Requirements i

i In accordance with the guidance in NUREG-1434, the licensee has proposed to 1

relocate portions of existing TS 3.10.3 " Shutdown Margin Demonstrations" within the improved TS.

Existing TS 3.10.3.a. and 4.10.3.a state the operability and surveillance testing requirements for SRMs in Mode 5.

The j

improved TS address these requirements.within specification 3.3.1.2 " Source

?

Range Monitor Instrumentation." Thus the SRM operability requirements in the l

existing TS for Mode 5 are effectively retained.within the improved TS. This i

change is considered an administrative change in the location of the i

requirements within the TS and is acceptable.

)

C.

More Restrictive Requirements l

j By electing to implement the NUREG-1434 Section 3.10 Specifications, the licensee has accepted a number of more restrictive conditions than are j

required by the existing TS. The more significant conditions are the following:

1.

The frequencies for existing TS 4.10.2.a have been changed from "Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to" and "once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />" to continuously during movement of control rods.

It is more conservative and appropriate to verify control rod movement at all times when the control rods are bypassed in the rod pattern control system (RPCS).

2.

Reactor mode switch interlock testing was incorporated as a note in-existing TS Table 1.2 " Operational Conditions," but no actions or surveillances were provided. Appropriate actions and surveillances are provided in improved TS 3.10.2 for this condition, which makes the-improved TS more restrictive and conservative in this area.

3.

Improved TS 3.10.3 is based on the allowance to withdraw a single control rod while in a shutdown mode from existing TS Table 1.2 footnote ***.

l However, improved TS 3.10.3 has an additional restriction applied. The existing TS has no specific requirement for this control rod to be capable of scram insertion (control rod operability and control rod drive accumulator LCOs are not applicable) to protect the core from the consequences of an inadvertent reactivity excursion.

Furthermore, the reactor protection system (RPS) requirements do not currently require the trip on Scram Discharge Volume (SDV).during this condition. The improved TS incorporates additional restrictions to address these issues. The i

option is provided to have an operable RPS SDV trip and an operable control rod (improved TS LCO 3.10.3 item d.1), or to appropriately preclude the possibility of a local reactivity excursion (improved TS LCO 3.10.3 item d.2).

The administrative controls required in this latter option are those currently licensed in existing TS 3.9.10.1.c and d for similar operations in the refuel mode.

Furthermore, actions and surveillances are also provided. The added actions will ensure appropriate operator response in the event one or more requirements are not met, and specific surveillances will ensure appropriate periodic confirmation of the required controls.

I 1

i

j.-

i j

i i

149 4.

Improved TS 3.10.4 " Single Control Rod Withdrawal - Cold Shutdown" separates the action associated with existing TS 3.9.10.1 " Single Control Rod Removal" into two conditions dependent on whether the affected control rod is insertable or not.

Condition A is a more detailed presentation of the existing requirement to " initiate action to satisfy the above requirements" and is considered an administrative change. By i

virtue of knowing that the control rod is insertable, a more explicit and j

restrictive action can be given. This action is to return the reactor i

mode switch to " Shutdown," which will preclude withdrawal of any control i

rod. This action will result in exiting the Applicability of the Special' i

1 Operation LCO and returning the reactor mode switch to its required i

position for normal Mode 4 operation, i

5.

Improved TS 3.10.5 " Single Control Rod Drive Removal - Refueling" adds additional requirements to the requirements specified in existing TS i

3.9.10.1 for this condition. These requirements are:

(a) a control rod withdrawal block is inserted and (b) no other Core Alterations are in progress. These requirements ensure no inadvertent criticality can occur, and represent an additional restriction on plant operations. New 1

surveillances have been added to verify a. control rod withdrawal block is inserted every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and to verify no other Core Alterations are in progress every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. These surveillance requirements ensure the requirements of the LCO are met. This change represents additional-restrictions on plant operations.

i

)

6.

The actions associated with existing TS 3.9.10.2 require that all i

positive reactivity changes be suspended if non-compliance with the LCO requirements occurs. The actions provided in the existing TS would not i

prohibit continued fuel loading.

Improved TS 3.10.6 has.an additional-required action which would suspend fuel loading upon non-compliance with the LC0 requirements. This additional restriction results in an j

appropriately conservative action for this LCO condition.

The staff has reviewed the above more restrictive generic requirements and concludes they result in an enhancement to the improved TS. Therefore, the j

more restrictive requirements are acceptable.

D.

Less Restrictive Requirements By electing to implement the NUREG-1434 Section 3.10 specifications the licensee has adopted a number of less restrictive conditions than are allowed j

by the existing TS. The more significant conditions are the following:

4 1.

Improved TS 3.10.1 " Inservice Leak and Hydrostatic Testing" has been added to allow performance of an inservice leak or hydrostatic test j

during Mode 4 with average reactor coolant temperature greater than 200*F (normally corresponding to Mode 3). With increasing reactor vessel fluence over time, hydrostatic and inservice leak testing will eventually be required to be performed with minimum reactor coolant temperatures greater.than 200*F. This would result in requiring the tests to be performed during Mode 3 operation. During Mode 3, the S/RVs are required

)

to be operable; however, performance of the hydrostatic tests-requires i

2

h i

i 1

y 150 the S/RVs to be gagged.

In addition, numerous other Mode 3 LCOs would also be applicable.

Because of the low decay heat levels during the hydrostatic test and the nearly water solid condition of the reactor i

pressure vessel, the stored. energy in the. reactor core is very low.

For the purposes of this test, the protection provided by normally required Mode 4 applicable LCOs, in addition to the specified Mode 3 applicable LCOs that are required to be met, will ensure acceptable consequences j

during normal hydrostatic test conditions and during postulated accident conditions.

)

i 2.

Improved TS 3.10.2 ' allows reactor mode switch interlock testing to be performed in Modes.3, 4, and 5 with the reactor mode switch in the Run,.

l l

Startup or Refuel position..The existing TS' require that-all control rods be fully inserted to perform the interlock tests.. Existing.TS j

3.9.10.2 provides an allowance for additional reactivity insertions l

(control rod removal) if all fuel assemblies in the control cell are j

removed. With one or more cells in this configuration the overall SDM is greater than when all control rods and fuel assemblies are inserted.

This same rationale can be applied for reactor mode switch interlock 2

testing.

Improved TS 3.10.2 incorporates this relaxation; however, an additional restriction of "no core alterations are in progress" has been included to prevent additional positive reactivity insertions from occurring.

i 3.

Existing TS 3.9.10.1 specifies certain requirements that must be met when i

removing a control rod in Mode 4.

Alternative requirements have been provided in improved TS 3.10.4 for Mode 4 operation in place of the SDM and control rod five-by-five array disarming requirements.

The 5

alternatives require all Mode 5 RPS functions to be operable and improved TS 3.9.5, Control Rod Operability - Refueling, to be made applicable.

These requirements ensure that if an inadvertent criticality occurs, the

)

RPS will initiate a scram and the withdrawn control rods will insert. A j

new surveillance has-also been added to verify every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that a j

control rod withdrawal block is inserted (SR 3.10.4.4) if the block is the chosen alternative requirement.

I 4.

The frequencies specified in existing TS 4.10.3, 4.9.10.1 and 4.9.10.2.1 have been changed for some of the components and systems specified in the i

surveillances. Most of the frequency intervals have been increased, however, a few have been decreased. The frequency intervals are consistent with the intervals specified in the improved TS for those components and systems when they are. required to be operable in Modes 1, 2 and 3, and they are deemed adequate once the test has begun to assure i

the necessary equipment operability and to assure other controis are met.

The additional burden posed by requiring a more frequent interval prior r

to commencing the test is not warranted.

If the surveillance has not i

been satisfactorily performed with the interval specified in the applicable SRs the test can not begin. This ensures the requirements are

}

adequately checked prior to and during performance of the tests and/or i

operations specified in the improved TS.

i

151 5.

The restrictions in existing TS 3.9.10.2 on fuel loading with control rods withdrawn have been relaxed in improved TS 3.10.6 to allow loading in a spiral reload sequence. Normal refueling procedures involve the' replacement of some fuel assemblies with new fuel, and rearrangement of the remaining fuel. Often this activity is performed in a shuffle sequence involving a scattering of only a few assemblies removed from-the core at any one time.

In this procedure all control rods remain fully:

inserted during all fuel loading operations.

On occasion'it is desirable'to off-load the entire core and then reload j

it with the new configuration.

In this event a significant number of i

blade guides would be required (two for each control cell). Alternate i

procedures have been developed which allow control rods to remain l

withdrawn during the core reloading process provided a specific' spiral reload procedure is followed. This spiral reload ensures:-

l a.

The control cell being reloaded has its control rod inserted;-

i b.

all other control cells, with one or more fuel assemblies inserted, have the associated control rod inserted; c.

the region of the core containing fuel (and associated inserted control rods) is contiguous, and contains at least one operable SRM; and d.

in the re'gion of the core containing no fuel assemblies, only control cells located immediately adjacent to the fueled region are loaded (after its control rod is inserted, see item a above).

This procedure assures that the spiral sequence is implemented when multiple control rods are withdrawn during refueling.

j The above less restrictive requirements have been reviewed by the staff and have been found acceptable because they do not present a significant safety question in the operation of the plant. The TS requirements that remain are consistent with current licensing practices, operating experience, and plant i

accident and transient analyses, and provide reasonable assurance that the l

public health and safety will be protected.

4 f

4.0 Design Features i

l This section contains the same material as found in the existing technical specifications (TS) except for those less restrictive specification-changes i

adopting NUREG-1434, which if altered in accordance with 10 CFR 50.59, would 1-not result in a significant impact on safety (the criteria of 10 CFR 50.36(c)(4) for including an item in the TS as a design feature).

(

4 i

j 1

i 2.

1 A

. +. - -

e.. - - -, -, -

m.-

+n.3,

.e

...E..,

e-.

.y me....

.e.

-., + -. -

p',. -. - -

,-e-4 1

t-1 I

l 152 A.

Significant Administrative Changes In accordance with the guidance in the Final Policy Statement, tha licensee i

has proposed administrative changes to the existing TS to bring them into j

conformance with the improved TS. These changes are as follows:

1.

The licensee has made an administrative change to the existing TS 5.1.

]

The existing TS includes figures that show maps of the GGNS site and 2

surrounding areas giving details of the exclusion area boundary and the 2

site boundary.

In place of the figures, improved TS 4.1 includes i

i description of the site location and exclusion area boundary. The descriptive requirements result in the same limits as the current i

requirements, because they represent an equivalent presentation of the existing TS intent. The improved TS changes are purely administrative and are acceptable.

?

l 2.

The licensee proposed to alter the discussion of the construction details

{

of existing TS 5.3 that require fuel rod cladding material be constructed of zircaloy alloy. The change is purely administrative in that it adopts the ZIRLO cladding design option allowed by 10 CFR 50.46(a)(1)(1), for nuclear power reactors fueled with uranium dioxide pellets.

The staff concludes that these provisions are not required to be in the TS i

j under 10 CFR 50.36(c)(4).

Further, the changes are purely administrative, and are therefore acceptable.

1 8.

Relocated Requirements a

j In accordance with the guidance in NUREG-1434, the licensee has proposed to relocate or reorganize all or portions of the following existing TS to other 3

licensee-controlled documents:

}

Existina TS Title 3

5.1.1 Exclusion Area l

5.1.2 Low Population Zone 5.1.3 Unrestricted Area and Site Boundary for Gaseous Effluents and for Liquid Effluents

]

5.2.1 Containment Configuration i

5.2.2 Design Temperature and Pressure 5.2.3 Secondary Containment j

5.3.1 Fuel Assemblies j

5.3.2 Control Rod Assemblies 5.4.1 Reactor Coolant System Design Temperature and i

Pressure i

5.4.2 Reactor Coolant System Volume i

5.5.1 Meteorological Tower Location 1

5.6.3' Fuel Storage Capacity 5.7.1 Component Cyclic or Transient Limit ia 4

i

e ja i

153 The more significant changes resulting from relocated items are as follows:

1.

Configurations, design temperatures and pressures, and volumes of Primary Containment and Drywell, Secondary Containment, and the Reactor Coolant i

System remain detailed in UFSAR-Sections 6.2.1, 6'2.3, and 5.1 and 5.2, i

respectively. Changes to these facility design parameters are controlled i

by the requirements of 10 CFR 50.59.

Furthermore, these design parameters are related to existing TS Limiting Condition for Operations (LCOs) that establish acceptable requirements for ensuring that performance of the containment and reactor coolant system is maintained i

and that any changes which may impact saft:ty would receive prior NRC review and approval. Since the features with a potential to impact safety are sufficiently addressed by LCOs, and since design features, if altered in accordance with 10 CFR 50.59, would not result in a i

significant impact on safety, the criteria of 10 CFR 50.36(c)(4) for i

including the above design features in the TS are not met.

3 2.

The figure identifying the location of the site meteorological tower in l

existing TS 5.5 is relocated to the UFSAR. The relocation of this l

location identification does not affect plant safety.

l 3.

Design feature specifications in the Component Cyclic or Transient Limits Table 5.7.1-1 are relocated from this section to other licensee-i

^

controlled documents. The licensee has adopted NUREG-1434 Section 4.0 i

and Section 5.5.5 as they relate to these changes. The table will be relocated to UFSAR Table 3.9-35 and the new program requirements, l

Specification 5.5.5 " Component Cyclic and Transient Limit," will be i

implemented to provide controls to track reactor pressure vessel cyclic and transient occurrences, to ensure that components are maintained-within the design limits. The staff finds the NUREG-1434 change together

)

with relocating the Table to the UFSAR will provide a level of protection and control equivalent to the existing plant specifications. Therefore, i

i j

the control of Component Cyclic and Transient Limits under 10 CFR 50.59 and the provisions of the administrative controls program are acceptable.

1 f

4.

Existing TS 5.1 includes figures that show maps of the GGNS site and surrounding areas giving details of the exclusion area boundary, the low i

population zone and the site boundary.

In place of the figures, improved TS 4.1 includes a description of the site location and the exclusion area i

boundary.

?

l The specific boundary for the Unrestricted Area remains detailed in UFSAR

]

Section 2.1.1.3.

The requirements for and restrictions on locating the i

Unrestricted Area must conform to regulations found in 10 CFR Part 20.

The low population zone is determined in accordance with-10 CFR Part 100 and this zone is identified and discussed in the UFSAR.

Evaluation of 5

changes to this feature of the facility is required in accordance with 10 CFR 50.59, and 10 CFR Part 100, as applicable.

Relocation of existing TS 5.1 design feature figure, showing the location of the low population I

i zone boundary, will not result in a significant impact on the safe operation of the facility because the UFSAR descriptions will continue to i-i

-n.

., - - - + -

-.---~w

-er a

.--e-,.,

o I

I 154 provide the information necessary to establish _the appropriate limits l

required by 10 CFR Part 100.

i l

5.

Existing TS 5.3.1 design features for the no'ainal active fuel length, and i

the specific design nominal enrichment of fuel assemblies remain detailed j

in UFSAR Section 4.2 and in other licensee-controlled documents. The-construction of the tubes containing the boron carbide, the nominal i

length containing the boron carbide, and the consistency of the boron carbide (i.e., " powder") of existing TS 5.3.2 remain detailed in UFSAR i

Section 4.2.2.4 and Figure 4.2-6b.

Changes to these facility design parameters are controlled by the requirements of 10 CFR 50.59..

i Furthermore,_ these design parameters are related to existing TS Limiting :

4 Condition for Operations (LCOs) for shutdown margin and thermal limits i

that establish acceptable requirements for ensuring that performance of fuel assemblies is maintained and that ar4y changes which may impact j

safety would receive prior Nuclear Regulatory Commission review and approval. Since the features with a potential to impact safety are sufficiently addressed by LCOs, and since design features,.if altered in t

accordance with 10 CFR 50.59, would not result in a significant impact on safety, the criteria of 10 CFR 50.36(c)(4) for including the above desiga p

features-in the TS are not met.

6.

Administrative limitations of existing TS 5.6.3 for spent fuel storage pool design requirements are relocated to the UFSAR Section 9.1.2.2.1.

i This limitation is not related to materials of construction or geometric j

arrangements that, if altered or modified, would have a significant i

impact on safety. The improved TS will retain sufficient information concerning design of the spent fuel pool consistent with 50.36(c)(4) and the Atomic Energy Act of 1954 as amended (the Act.) Therefore, requirements for design features are not met and the existing TS 5.6.3 requirements are relocated to the UFSAR.

j i

j The above relocated requirements relating to design features are not required i

to be in the TS under 10 CFR 50.36(c)(4) or the Act and are not required to j

obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety.

In addition, the staff

?

finds that' sufficient regulatory controls exist under 10 CFR 50.59.

Accordingly, the staff has concluded that these requirements may be relocated l

from the TS to the licensee's TS Bases or to the UFSAR, as applicable.

1 l

C.

More Restrictive Requirements 8I By electing to implement the NUREG-1434 Section 4.0 Specifications, the j

licensee has adopted more restrictive conditions than are required by the existing TS. The existing TS do not contain k-effective and nominal-fuel 4

assembly-storage spacing limitations on fuel storage in the new fuel storage racks. The addition of these specifications in the improved TS clarify current plant practices that assure the fuel storage pools-remain subcritical for design basis fuel handling accidents. The staff has reviewed the more restrictive requirements and concludes that they result in an enhancement to j-the improved TS.

~.

l 155 D.

Less Restrictive Requirements The licensee in electing to implement NUREG-1434 Section 4.0 specifications has proposed a.less restrictive condition that eliminates a reference to limit the number of lead test assemblies that.are placed in non-limiting core regions than is allowed by their existing TS. This allowance provides a recognition of a specific kind of special test that may be perfomed. This is intended to avoid potential confusion regarding whether a TS change is required to conduct this tert. The requirements of 10 CFR 50.59 regarding conducting special tests remain applicable. Since the features with a l

potential to impact safety are sufficiently addressed by LCOs, and since design features, if altered in accordance with 10.CFR 50.59, would not result in a significant impact on safety, the criteria of 10 CFR 50.36(c)(4) for including the above design features in the TS are not met.

The above less restrictive requirements have been reviewed by the staff and have been found to be acceptable, because they do not present a significant safety question in the operation of the plant.

The TS requirements that remain are consistent with current licensing practices, operating experience and plant accident and transient analyses, and provide reasonable assurance that the public health and safety will be protected.

5.0 Administrative Controls The licensee has adopted the NUREG-1434 Specifications for Section 5.0 with l

some plant specific differences due to current licensing basis.. The licensee t

has modified improved TS 5.6.6 reactor coolant system (RCS) Pressure and Temperature Limits Report (PTLR), to reflect the return of this information to the appropriate improved TS limiting condition for operations.

i A.

Significant Administrative Changes In accordance with the guidance in the Final Policy Statement, the licensee has proposed administrative changes to the existing TS to bring them into conformance with the improved TS. These changes are as follows:

1.

The technical contents of several existing TS requirements are being relocated from other existing TS chapters to the administrative control section. These requirements are to become programs in the improved TS.

A statement of applicability of improved TS SR 3.0.2 or SR 3.0.3 is needed in those new programs to maintain the current allowances for sur-veillance frequency extensions since these SRs are not normally applied to frequencies identified in the administrative controls section of the TS. The following improved TS programs incorporate this change:

Existino TS Title 5.5.6 Inservice Testing Program 5.5.7 Ventilation Filter Testing Program 5.5.8 Explosive Gas and Storage Tank Radioactivity Monitoring Program t

I 4

156 4

Since this change maintains current requirements within the improved TS, it is considered an administrative change and is acceptable.

2.

The requirements, qualifications and duties of the Shift Technical Advisor (STA) are specified in a number of specifications-in existing TS 6.2.

Improved TS 5.2.2.g combines all of these items into one 1

specification by identifying the duties of the STA and referring to the 4

Commission's Policy Statement on STA and the procedures that implement the policy statement to define the other characteristics and details 1

associated with the STA. The entire change is considered an administrative change and is therefore acceptable.

3.

Existing TS 6.9.1.4 " Annual Reports" requires that the initial annual reports be submitted prior to March 1 of the year following initial criticality.

Since the plant is an operating plant, the initial report has already been submitted. Therefore, the requirement is no longer applicable and its deletion from the improved TS is an acceptable administrative change.

4.

Existing TS 6.9.1.5.2 requires an annual report on challenges to the safety valves and safety / relief valves.

Existing TS 6.9.1.10 requires the same information on a monthly basis. The improved TS deletes the annual report requirement but retains the monthly reporting requirement on the valves. Since no change in the details of the reporting are required, this is only a change in when the report is to be submitted and j

thus, is considered an administrative change and is acceptable.

5.

Existing TS 6.12.2 and 6.12.3 define high radiation areas in terms of radiation levels that could result in a specified theoretical dose to an individual in one hour.

It is not necessary to specify this number since i

the maximum permissible occupational dose is specified in 10 CFR Part 20.

Therefore, this limit is deleted from the improved TS. This is an i

administrative change which is acceptable.

1 6.

NUREG-1434 Specification 5.5, Explosive Gas and Storage Tank Radioactivity Monitoring Program, specifies the requirements and limits for the main condenser offgas treatment system and unprotected outdoor liquid radwaste tanks. The licensee in improved TS 5.5.8 has modified 1

the requirements by deleting the references and requirements which do not I

apply to the current plant design.

In addition NUREG-1434 Specification 5.5.8.c has been modified in improved TS 5.5.8.b to reflect the existing TS requirements of existing TS 3/4.11.1.4, Liquid Holdup Tanks.

Since i

these requirements result in the same limits as the current requirements, the changes are purely administrative and are therefore acceptable.

1 7.

NUREG-1434 Specification 5.5.9, Diesel Fuel Oil Testing Program, specifies the requirements, standards and limits to be followed in testing and maintaining quality diesel generator fuel oil. The licensee in improved TS 5.5.9 has modified the requirements, standards and limits to reflect the existing TS requirements of existing TS 4.8.1.1.2.d.

4 4

4 H

l

]=

157 Since these requirements result in the same limits as the current requirements, the changes are purely administrative and are therefore acceptable..

5 8.

NUREG-1434 Specification 5.9.1.7, RCS PTLR, specifies the analytical i

methods to be used to determine the RCS pressure and temperature limits l

and specifies that these limits for the reactor pressure vessel and RCS i

shall be documented in a report maintained outside of TS. The licensee i

has elected to retain this information within the TS. Thus, the pressure l

and temperature limits and curves specified in existing TS Section 3/4.4 will be included in the appropriate specifications of improved TS Section j

3.4, Reactor Coolant System. Since the current requirements are retained, the improved TS is acceptable.

l i

9.

NUREG-1434 Specification 5.7, High Radiation Area', specifies the requirements that may be used in lieu of. the requirements specified in 10 i

CFR 20.1601(a) as allowed by 10 CFR 20.1601(c). The licensee has l

modified these requirements in improved.TS 5.7 to reflect the i

requirements of existing TS 6.12, High Radiation Area. Since these j

requirements result in the same limits as the current requirements, the j

changes are purely administrative and.are therefore. acceptable.

10. A footnote to existing TS 6.9.1.4, " Annual Reports," states that a single report may be submitted for a multi-unit facility. The note was inserted i

into the existing TS on the assumption that-Grand Gulf Unit 2 would be i

completed. The licensee proposes deleting the note from the Grand Gulf

{

improved TS because Unit 2 has been cancelled. This administrative j

change is acceptable.

l B.

Relocated Requirements i

i In accordance with the guidance in NUREG-1434, the licensee has proposed to i

relocate or reorganize all or portions existing-TS 6.7, Safety Limit j

Violation, within the improved TS.

Existing TS 6.7 specifies the actions to 1

l be taken in the event a safety limit is violated. The improved TS addresses these items within Specification 2.2, SL Violations. Thus the safety limit i

violation actions in the existing TS are effectively retained within the j

improved TS. This change is considered an administrative change in the J.

location of the requirements within the TS and is,,therefore, acceptable.

i In accordance with the guidance in the Final Policy Statement, the licensee has proposed to relocate or reorganize all or portions of the following j

Existing TS to other licensee-controlled documents:

]

Existina TS Title 6.1.2 Line of Authority Directive i

6.2.1.a Organizational Structure j

6.2.2.a Minimum Shift Composition L

6.2.2.g Operating Licensing Requirements Table 6.2.2-1 Minimum Shift crew Composition

)

6.2.3 Independent Safety Engineering Group j

6.3 Unit Staff Qualifications 1

4 i

w

-we.,

e 9..

.-engg--.

yv,p-w en v

.-pty.

i9 i

j 158-l.

-6.4 Training i

6.5 Review and Audits l

6.6 Reportable Event Action 6.7 Safety Limit: Violation 6.8.1.f Fire Protection Program Implementation 6.8.1.g Process Control Program Implementation 6.8.1.1 Quality Assurance Program for Effluent and Environmental Monitoring 6.8.2 Review and Approval Process j

6.8.3.b In-Plant Radiation Monitoring Program 6.8.4.e Radiological Environmental Monitoring Program-i j

6.9.1.1 thru 6.9.1.3 Startup Reports I

6.9.1.11 Core Operating Limits Report

{

6.10.

Record Retention 6.11 Radiation Protection Program i

6.12 High Radiation Area 6.13 Process Control Program (PCP) 6.14 Offsite Dose Calculation Manual (ODCM) l The more significant changes resulting from relocated items are as follows:

j 1.

Line of Authority 1

i_

Existing TS 6.1.2 requires that an administrative letter be issued to station personnel on an annual basis describing responsibility to the j

j shift superintendent. This is unnecessary since the organization and j'

responsibilities of each function are adequately described in the licensee's UFSAR. Repeating the organization responsibilities via an internal management directive only increases the administrative burden on j'

the facility with no resulting benefit.

In addition plant safety is not compromised by this proposed deletion from the TS. Therefore, control of i

these provisions under 10 CFR 50.59 is acceptable. The staff concludes that the regulatory requirements provide sufficient control of this i

provision and removing it from the improved TS is acceptable.

l 2.

Minimum Shift Crew Composition 3

The licensee proposes that existing TS 6.2.2.a and associated Table 6.2.2-1 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, 2

as well as for cold shutdown and refueling. The requirements in this

{

Specification and the associated table are located in the UFSAR.

i Additionally, improved TS 5.1.2,. 5.2.2.a, 5.2.2.b, and 5.2.2.c 'specify

{

when licensed and non-licensed operators 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 the j

TS is acceptable.

i i

i.

L a

1 17 l

j 159 l

1 l

3.

SRO Present Durino Fuel Movement j

The licensee proposes that the requirement in existing TS 6.2.2.d that an SRO be present during fuel handling and supervise all core alterations 4 -

not be retained in the TS. The SR0's presence in these conditions is required by 10 CFR 50.54(m)(2)(iv) and need not be controlled by TS to

]

]

assure safe operation of the facility. The current regulation states:

j "Each licensee shall have present, during alteration of the core of a j,

nuclear power unit (including fuel loading or transfer),.a person-i holding a senior operator license or a senior operator license i

limited to fuel handling to directly supervise the activity and,-

j during this time, the licensee shall not assign other duties to this

-i

' person."

This requirement is specified in. plant procedures which implement 10 CFR 50.54. The staff concludes that the regulatory requirements provide 5

i sufficient control of these provisions and removing them from the TS is i

j acceptable.

1 j

4.

10flependADt Safety Enaineerina Groun 4

The requirements of existing TS 6.2.3 on the Independent Safety.

j Engineering Group-(ISEG) may be deleted from the.TS on the basis that-they are adequately addressed by improved TS Section 5.0 administrative l

l controls as well as regulations.

Improved TS 5.3, Unit Staff i

Qualifications, provides adequate requirements-to assure an _ acceptable, competent unit staff. Each member of the unit staff shall'aeet or exceed the minimum qualifications of specific Regulatory Guides or ANSI Standards acceptable to the staff.

Improved TS 5.3 describes the details of the required qualifications. Additionally, improved TS 5.2, o

Organization,, details unit staff requirements.

Further, the UFSAR l

describes compliance with NUREG-0737, including the requirements for an ISEG.

}

The licensee has proposed to relocate these provisions to the UFSAR, as previously described. The staff concludes that the control of these 3

provisions under 10 CFR 50.59 and 10 CFR 50.54(m) is sufficient and-

{

removing them from the TS is acceptable.

7-5.

Trainina j

i The licensee proposes that the requirements of existing TS 6.4 on i

training be relocated to the UFSAR. Training and requalification of.

1 those positions are as specified in 10 CFR Part 55.

The licensee has proposed to relocate these provisions to the UFSAR, as previously described. The staff concludes that the control of these provisions j

under 10 CFR 50.59 is sufficient control and removing them from the TS is j

acceptable.

4 b-i

.m

~., _ _

P e

160 1

1 6.

Review and Audits 1

i The licensee proposes that the review and audit functions specified in

)

j existing TS 6.5, and 6.6.1.b be relocated from the improved TS on the i

basis that they are adequately controlled elsewhere.. These TS provisions 1

are not necessary to assure safe operation of the facility, given the requirements in the Quality Assurance (QA) Program implementing 10 CFR 50.54 and 10 CFR Part 50, Appendix B to control the' requirements for all i

review and audit functions except those associated with the security and emergency plans. The security and emergency plan review and audit l

functions are relocated to their respective plans in accordance with GL i

93-07. Such an approach would result in an equivalent level of regulatory authority while providing for a more appropriate change control process. The level.of safety of plant operation is unaffected by this change and NRC and licensee resources associated.with processing i

license amendments to this administrative control may be used more

]

effectively.

In addition, the following considerations support relocating these items from the TS:

c.

The on-site review function, composition, alternate membership, j-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 description and change control is j

provided by 10 CFR 50.54(a).-

i b.

The off-site review group is also addressed, although with less

]

detail, in ANSl'N18.7-1976. The'QA program description' include the requirements for the off-site review group. Therefore, duplicating the review and audit function of the off-site review group in the improved TS is unnecessary.

j 1

1 Audit requirements are specified in the QA program description to c.

satisfy 10 CFR Part.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 Part 73. Therefore, duplication of these requirements i

does not enhance the level of safety of the plant, nor are the l

provisions relating to audits necessary to assure safe operation of the facility.

The licensee has proposed to relocate the provisions that are not i

otherwise covered by regulatory requirements to the UFSAR. The staff j-concludes that the sufficient regulatory controls exist for the UFSAR i.

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

the UFSAR under the controls of 10 CFR 50.54.is acceptable.

I 7.

Reportable Event Action

[

The licensee proposes that the requirement in existing TS 6.6.1.a that the Commission be notified of all reportable events not be retained in the TS.

10 CFR 50.73(a)(2) provides requirements for the licensee to i

submit a Licensee Event Report (LER) for all reportable events'specified in 10 CFR 50.73. The reports are required to be submitted within 30 days i

4 h

.. _. _.. _, _ ~ _ _ _, _,.,,

=-.

l 161 and will contain the same type of information required by existing TS j

6.6.1.a.

The above requirements are included in licensee procedures which implement 10 CFR 50.72 and 10 CFR 50.73. The staff concludes that these regulatory requirements provide sufficient control of these

]

provisions and removing them from the TS is acceptable.

I l

8.

Safety Limit Violation i

The licensee proposes that the requirements in existing TS 6.7.1.b for i

safety limit violation report content and the time for providing this report to utility management not be retained in TS. The general requirements for the report's conte.nts are dictated by 10 CFR 50.73 for i

LERs. These requirements are included in licensee procedures which implement 10 CFR 50.72 and 10 CFR 50.73. The requirement to provide the report to onsite safety review group specified in existing TS 6.7.1 are-i to be relocated to the UFSAR. Changes to the UFSAR will be controlled by 10 CFR 50.59. The staff concludes that the control of these provisions under 10 CFR 50.59 is acceptable, that the regulatory requirements provide sufficient control of these provisions and that removing them j

from the TS is acceptable.

H 9.

Security Plan and Emeraency Plan Imolementation l

1 The licensee proposes to relocate' the requirements to establish, imple-ment, and maintain procedures related to the Emergency Plan (existing TS 6.8.1.e) and Security Plan (existing TS 6.8.1.d).

Since the Security i

Plan requirements are specified in 10 CFR 50.54, 73.40, 73.55, and 73.56 l

and the Emergency Plan requirements are specified in 10 CFR 50.54(q) and 10 CFR Part 50, Appendix E, Section V, the staff in GL 93-07 removed the j

requirements from the STS and relocated them to their respective plans.

The requirements in the existing TS for the review of the security program and implementing procedures, and for the review of the station i

emergency plan and implementing procedures, will be included in the UFSAR.

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) i for the Emergency Plan.

]

The staff concludes that the extensive requirements for emergency planning in 10 CFR 50.47, 50.54, 10 CFR Part 50 Appendix E 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 i

the programs will be met. Therefore, duplication of the requirements i

contained in the regulations would not enhance the level of safety for l

the facility. The staff concludes that other regulatory requirements i

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

I-

~-.-.- -

4 i-1 162 l

10. Process Control Proaram The licensee proposes to relocate the implementation requirements of j

j-

-existing TS 6.8.1.g and the program description in existing TS 6.13 for i

the Process Control Program (PCP). The PCP is described in the UFSAR.

The PCP implements the requirements of 10 CFR Part 20, Part 61, and Part

71. Relocating the description of the PCP does not affect the safe operation of the facility. The' staff concludes that the. regulatory con-trols for the UFSAR provide sufficient control of these requirements and j

removing these provisions from the TS is. acceptable, i

11. Review and Anoroval Process The licensee is proposing to relocate the requirements of existing TS' L

6.8.2 for the review and approval process for procedures to the UFSAR.

1 i

This proposal is based on the existence of the following requirements j

i which duplicate 10 CFR Part 50 Appendix B.in these areas.

i t

l The requirement for procedure control is mandated by 10 CFR.Part 50, I

Appendix B, Criterion II and. Criterion V.

ANSI N18.7-1976, which is

')

endorsed by the NRC in Regulatory Guide1.33, " Quality Assurance Program 1

l Requirements (Operations)," is used in the development of many licensee I

4 QA program. descriptions, and contains specific requirements related to procedures. The licensee has committed to follow ANSI N18.7-1976 as a l'

means to comply with 10 CFR Part 50, Appendix 8.

ANSI N18.7-1976, i

Section 5.2.2 discusses procedure-adherence; Section 5.2.2 clearly states that procedures shall be followed, and the requirements for use of j

procedures shall be prescribed in writing. ANSI N18.7-1976 also requires review and approval of procedures to be defined. ANSI N18.7-1976, j

Section 5.2.15 describes the review, approval and control of procedures.

j The ANSI standard describes the requirements to provide measures to control and coordinate the approval and issuance of documents, including i

changes thereto, which prescribe all activities affecting quality. - The ANSI standard further states that each procedure shall be reviewed and i

approved prior to initial use and describes the required reviews. ANSI L

N45.2-1971, Section 6, also specifies that the QA Program describe i

procedure requirements.

The licensee will continue to implement a QA program description in accordance with the requirements of 10 CFR Part 50, Appendix B, which i

provides appropriate controls for the review and approval of procedure i

changes. Changes to the QA program description that is incorporated in 3

the UFSAR by reference, including departures from the ANSI standard that constitute a reduction in commitment, will be governed by 10 CFR

]

50.54(a). The staff concludes that these regulatory requirements provide sufficient control of these provisions and: removing them from the TS is acceptable.

j 7

12.

In Plant Radiation Monitorina The In Plant Radiation Monitoring Program (IPRMP) existing TS 6.8.3.b provides controls to ensure the capability to accurately determine the i

2 l

l

-_,_,J.

i*

I i

163 t

I i

airborne iodine concentration in vital areas under accident conditions.

i l

This program was developed to minimize radiation exposure to plant j

j personnel (post-accident) and is not relied upon to detect a degradation of the reactor coolant system pressure boundary. The licensee has i

j

. proposed to relocate these provisions to the.UFSAR and appropriate plant l

procedures. The staff concludes that the control of these provisions j

under 10 CFR 50.59 is acceptable.

l

13. Startun Renort i

The requirements in existing TS 6.9.1.1 through'6.9.1.3 to submit a Startup Report have been relocated to the UFSAR. The report was a l

summary of plant startup and power escalation testing following receipt i

of the Operating License, an increase in licensed power level, the j

installation of nuclear fuel with a different design or manufacturer than l

the current fuel, and modifications that may have significantly altered l

the nuclear, thermal, or hydraulic performance of the facility.

The report provides a mechanism for the staff to review the appropriateness i

of licensee activities after-the-fact, but contains no requirement for j.

staff approval.

Inasmuch as this report was required to be provided to i

j the staff within 90 days following completion of the respective i

j milestones, all of which have already occurred, the removal of this requirement is acceptable.

14. Soecific Activity Analysis Renort l

The licensee proposes that the requirement in existing TS 6.9.1.5.3 for.

I 1

the results of special activity analysis in which the primary coolant j

exceeds the limits of existing TS 3.4.5 be reported to the Commission, l

not be retained in the TS.

10 CFR 50.73(a) provides requirements for the l

l licensee to submit an LER to report fuel cladding failures that exceed i

i expected valves or that are caused by unexpected factors, e.g., where fuel cladding:is seriously degraded. The LERs will contain the same type 1

of information as that required by existing TS 6.9.1.5.3.

The above reporting requirements are included in the licensee procedures which 1

implement 10 CFR 50.72 and 10 CFR 50.73. The' staff concludes that these j

regulatory requirements provide sufficient control of these provisions j

and removing them from the TS is acceptable.

i l

15. Record Retention The licensee proposes that the requirements in existing.TS 6.10 on record retention be relocated from the improved TS on the basis that they are adequately addressed by the QA Program required by regulation (10 CFR Part 50, Appendix B, Criteria' XVII) and the UFSAR.

Facility operations are performed in accordance with approved written procedures. Areas include normal startup, operation and shutdown, abnor-mal conditions and emergencies, refueling, safety-related maintenance, i

surveillance and testing, and radiation control. Facility records docu-ment appropriate station operations and activities.

Retention of these records provides documentation retrieveability for review of compliance

Id

?

o i

i 164 with requirements and regulations.

Post-compliance review of records does not directly assure operation of.the facility in a safe manner, as-j i

activities described in these documents have already been performed.

In addition, numerous other regulations such as 10 CFR Part 20,' Subpart L, and 10 CFR 50.71 require the retention oF certain records related to 1

j operation of the nuclear plant. The staff concludes that these regula-i tory requirements provide sufficient control of these recordkeeping l

provisions and removing the requirements from the TS is acceptable.

.l l

l l

16. Radiation Protection'Prooram i

i i

l The licensee proposes to relocate the program description in existing TS l

6.11 for the Radiation. Protection Program to the UFSAR.- The Radiation l

2 l

Protection Program requires procedures to be prepared for personnel l

radiation protection consistent with the requirements of 10 CFR Part 20.

The requirement to have procedures to implement Part 20 is also contained.

i l

within 10 CFR 20.1101(b). Periodic review of these procedures is addressed under 10 CFR 20.1101(c). The program requirements specified l

above are described in the UFSA' l

The staff concludes that the requirements of the rule provide sufficient i

i control of these provisions, and that 10 CFR 50.59 provides adequate l

j controls for changes to those provisions in the UFSAR. Accordingly-i j

their relocation from the TS is acceptable.

1

17. Temoorary Chanae Process l

l The licensee proposes to relocate the requirements of existing TS 6.8.3 l

for the temporary change process for procedures to the UFSAR. This-4-

proposal is based on the existence of the following requirements which duplicate 10 CFR Part 50, Appendix B in these areas.

The requirement for procedure control is mandated by 10 CFR Part 50, i

i Appendix B, Criterion II and Criterion V.

ANSI N18.7-1976, which is

{

l endorsed by NRC in Regulatory Guide 1.33, " Quality Assurance Program i

Requirements (Operations)," is used in the development of many licensee j

QA program descriptions and contains specific requirements related to i

procedures. The licensee has committed to follow ANSI N18.7-1976 as a j

means to comply with 10 CFR Part 50, Appendix B.

ANSI N18.7-1976, Section 5.2.2 discusses procedure adherence. Section 5.2.2 clearly states that procedures shall be followed, and the requirements for use of 1

procedures shall be prescribed in writing. ANSI N18.7-1976 also discusses temporary changes to procedures. ANSI N18.7-1976, Section i

5.2.15 describes the review, approval and control of procedures. The ANSI standard 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 ANSI standard further states that each procedure shall be reviewed and approved prior to initial use, and describes the required reviews. ANSI N45.2-1971, Section 6, also specifies that the QA Program describe procedure requirements.

-w n,--,.

7

o i

165 i

The licensee will continue to implement a QA program description in accordance with the requirements of 10 CFR Part 50, Appendix B, which provides appropriate controls for the review and approval of procedure

]

l changes. Changes to the QA program description that is incorporated in 1

the UFSM by reference, including departures from the ANSI standard that j.

constitute a reduction in commitment, will be governed by 10 CFR 50.54(a). The staff concludes that these regulatory requirements provide i

sufficient control of these provisions and removing them from TS is 4

i l

acceptable.

l

18. Radioloaical Environmental Monitorina Proaram.

The Radiological Environmental Monitoring Program '(existing TS 6.6.4.e) l requires monitoring the radiation and radionuclides in the environs of l

GGNS consistent with the guidance specified in 10 CFR Part 50, Appendix I.

This program ensures that radioactive effluents.are restricted to levels as low as reasonably achievable, and have no impact on plant l

i nuclear safety. The details and description of the program are already j

contained in the ODCM, as specified by existing TS 1.26 and improved TS 5.5.1.

The staff concludes that these regulatory requirements and the-j UFSAR provide sufficient control of these provisions and removing them j

from the TS is acceptable.

19. Offsite Dose Calculation Manual Record Retention il Existing TS 6.14.a specifies that all changes to the ODCM shall be documented and records of reviews performed shall be retained as required by existing TS 6.10.3.

The requirements of existing TS 6.10 have been

]

relocated as described in Section 5.0.B above. Since existing TS 6.14.a i

requires the ODCM records be retained in accordance with existing TS 6.10.3, the justification for relocating existing TS 6.10 also applies to this requirement in existing TS 6.14.a.

Therefore, the record retention j

requirements of existing TS 6.14.a can be relocated to the UFSAR.

4 I

20. Annual UFSAR Undate i

Existing TS 6.2.1.a requires the licensee tu annually revise the i

organization charts and the lines of authority, responsibility, and j

communication depicted in the UFSAR. UFSAR updates are governed by 10 CFR 50.71(e)(4). This requirement is included in the licensee procedures which implement 10 CFR 50.71. Therefore, the organization structure-requirements of existing TS 6.2.1.a can be relocated to UFSAR.

j

21. Doerator Licensino Reauirements 1

Existing TS 6.2.2.g specifies each licensee staff position that requires a senior reactor operator license.

Improved TS 5.2, " Organization,"

specifies the licensee's organizational requirements. Specifications 5.2.2.a. 5.2.2.b, 5.2.2.c, and 10 CFR 50.54 describe the minimum shift i

crew composition and delineate which positions require an R0 or SRO license. Training and requalification of those positions are as

)

specified in 10 CFR Part 55.

i i

i i

i

m 1

i 166 i

i 1

l The licensee proposes to relocate these provisions to the UFSAR, as previously described. These provisions do not need to be controlled by I

TS under the Commission's regulations.

Therefore, the control of these t

j provisions under 10 CFR 50.59 is acceptable.

j

22. Trainina Proar =

The licensee proposes relocating existing TS 6.2.3, " Independent Safety Engineering Group (ISEG)," and 6.4, " Training," as described in paragraph 4.

The justifications for relocating existing TS 6.2.3 and 6.4 also j

apply to the training criteria in existing TS 6.3 because existing TS i'

6.3, " Unit Qualifications," specifies additional training criteria for Therefore, these items can be licensed personnel and ISEG personnel.

relocated to UFSAR.

i

23. Fire Protection Proaram Imolementation Existing TS 6.8.1.f specifies that the procedures for the Fire Protection l

Program shall-be established, implemented, and maintained. The Fire I

Protection Program includes controls to ensure that appropriate fire protection measures are maintained to protect the plant from fire and tox ensure the capability to achieve and maintain safe shutdown in the event' l

j of a fire. The program was originally developed to ensure the capability

{

to provide for alternate and dedicated safe shutdown in accordance with j

Appendix R to 10 CFR Part 50. Therefore, it ensures the ability to place i

the unit in a' safe condition in the event of a fire..

i GL 86-10, " Implementation of Fire Protection Requirements," specified a l

stand 3rd license condition that required prior staff approval of changes to the Fire Protection Program that adversely affect the ability to achieve al.a maintain safe shutdown in the event of a fire. The license condition for Grand Gulf is license condition 2.c.(41) of Operating i

License NPF-29. The license condition also requires the implementation j

and maintenance of all provisions of the Fire Protection Program. Thus, j

the requirement of existing TS 6.8.1.f may be moved from the TS. The control of the Fire Protection Program and the implementation provisions i

under the license condition and 10 CFR 50.59 are acceptable.

}

24. Methods for Imolementina OA Procram for Effluent and Environmental 1

Monitorina

]

Existing TS 6.8.1.1 specifies the methods in the QA program for effluent i

and environmental monitoring. The methods to be used are dictated by the guidance in Regulatory Guide 4.15, the UFSAR, and plant procedures. The improved TS do not describe how to follow a specification; therefore, the methods and details can be moved from the TS to UFSAR. This change is l

acceptable because the methods and details can be controlled under 10 CFR j

50.59. The changes do not affect the requirements for compliance with 10

[

CFR Part 36a and 10 CFR Part 20.

i The above relocated requirements relating to design features are not required to be in the TS under 10 CFR 50.36, and are not required to obviate the 2

~

e 167 possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety.

In addition, the staff finds that sufficient regulatory controls exist under 10 CFR 50.59 and 10 CFR 50.54 and the other regulations set forth above to assure continued protection of the public health and safety. Accordingly, the staff has concluded'that these requirements may be relocated from the TS to the licensee's TS Bases, plant procedures or incorporated by reference in the UFSAR Chapte: 16, as applicable.

C.

More Restrictive Requirements 1

By electing to implement the NUREG-1434 Section 5.0 Specifications, the licensee has adopted a number of more restrictive conditions than are required by the existing TS. The more significant conditions are the following*

r 1.

The licensee has adopted the Safety Function Determination Program (improved TS 5.5.10) and Technical Specification Bases Control Program:

(improved TS 5.5.11), two more restrictive conditions than are required by the existing TS. The Safety Function Determination Program is 1

included to support implementation of the support system operability characteristics of the' improved TS. The Bases Control Program is provided to specifically delineate the appropriate methods and reviews necessary for a change to the improved TS Bases.

2.

The improved TS modify existing TS 6.9.1.7 to include the details of the information to be contained in the Annual Radiological Environmental i

Operating Report. The details were removed when the licensee implemented i

GL 89-01, but are being included in the improved TS because the details are necessary to assure the reports are provided with similar content and i

format for comparison with other plants and with prior reports.

3.

The licensee proposes a requirement in improved TS 5.4.1.b to maintain Emergency Operating Procedures (EOPs) as implemented in response to NUREG-0737. Although E0Ps are included as a necessary type of procedure in Regulatory Guide 1.33, the existing TS do not include the additional procedures and changes made in response to the guidance in NUREG-0737 and Supplement 1.

This change ensures these commitments made in response to GL 82-33 are maintained, and that the guidance and commitments are appropriately considered for any changes to these procedures.

The staff has reviewed the above more restrictive requirements.and concludes they result in an enhancement to the improved TS. Therefore, the more i

restrictive requirements are acceptable.

D.

Less Restrictive Requirements The licensee in electing to implement NUREG-1434, Section 5.0 Specifications has proposed a number of less restrictive conditions than are allowed by the existing TS. The most significant change is the requirement for the Manager-Plant Operations to have held an SR0 License and the requirement for the Operation Superintendent to hold an SRO License is replaced with a requirement I

that the operations manager or at least one operations middle manager hold an i

l m

l-l

)

]

168 SRO License. The less' restrictive requirement has been reviewed by the staff.

i-and has been found to be acceptable, and is consistent with the requirements 4

of Regulatory Guide 1.8,' Revision 2, and ANSI N18.1-1971. The TS requirements that remain are consistent with current licensing practices, operating i

experience and plant accident and transient analyses, and provide reasonable assurance that the public health and safety will be protected.

j IV - STATE CONSULTATION In accordance with the Commission's regulations, the Mississippi State i

official was notified of the proposed issuance of the amendment. The State official had no comments.

V ENVIRONMENTAL CONSIDERATION i

i Pursuant to 10 CFR 51.21, 51.32 and 51.35, an environmental assessment and finding of no significant impact has been prepared and published in the Federal Reaister on February 15, 1995 (60 FR 8739). Accordingly, based upon the environmental assessment, the staff has determined that the issuance of f

the amendment will not have a significant effect on the quality of the human i

environment.

l l

VI CONCLUSION The Grand Gulf Nuclear Station improved technical specifications (TS) provide clearer, more readily understandable requirements to ensure safe operation of l

the plant. The staff has concluded that the improved TS satisfy the guidance in the NRC Final Policy Statement with regard to the content of technical i

specifications, and conform to the model provided in NUREG-1434 with-appropriate modifications for plant-specific considerations. The staff has i

concluded that the Grand Gulf Nuclear Station improved technical'

]

specifications satisfy Section 182a of the Atomic Energy Act,10 CFR 50.36, and other applicable standards. On this basis, the staff concludes that i.he proposed Grand Gulf Nuclear Station improved technical specifications are acceptable.

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

Principal Contributors:

C. Schulten Date: February 21, 1995 i

i

. ~

f s

i I

i 169 i

4 Table 1 j.

Summary of Grand Gulf Nuclear Station 3

Totally and Partially Relocated Technical Specifications t

i i

Existing Title Relocation Relocation TS Document

' Control 2.1.4 Reactor vessel Water Level UFsAR/sases

$50.59 i

4.0.5 Innervice Inspection and Testins Programs UFSAR/sases

$50.55a 550.59 i

3/4.1.2 Reectivity Anannties UFSAR/sases

$50.59 i

q 3/4.1.3.1 control Rod OperabitIty UFSAR/ Bases

$50.59 3/4.1.3.2 centrol Rod Masieum scram Insertion Times UFsAR/sases 650.59 l

3/4.1.3.3 control Rod scram Accumulators UFSAR/sases

$50.59 3/4.1.3.4 control Rod Drive Co wline UFsAR/ Bases

$50.59 i

3/4.1.3.5 control Rod Position Indication UFSAR/ Bases

$50.59 3/4.1.3.6 centrol Rod Drlwe Housing steport UFsAR/saaes

$50.59 I

3/4.1.4.1 Control Rod Withdrawet UFsAR/ Bases

$50.59 d

1 l

3/4.1.4.2 Rod Pattern control system UFsAR/sases

$50.59

+

f 3/4.1.5 stan ty Liquid control system UFsAR/ Bases 550.59 Y

{

3/4.2 Power Distribution Limits UFSAR/sases

$50.59 3/4.3.1 Reactor Protection system Instrumentation UFSAR/ Bases

$50.59 4

f 3/4.3.2 Isoletion Actuation Instrumentation UFSAR/sases

$50.59 l

3/4.3.3 Emergency Core cooling system Actuation UFsAR/ sanes

$50.59 Instrumentation f

3/4.3.4.1 ATWs RPT system Instrumentation UFSAR/sases

$50.59 3/4.3.4.2 E0C RPT system Instrumentation UFsAR/sases

$50.59 l

j 3/4.3.5 RCIC system Instrumentation UFSAR/ Bases

$50.59 j

3/4.3.6 control Rod Block Insertamntation UFsAR/sases

$50.59

{

3/4.3.7.1 RadietIon Monitorins Instrumentation UFSAR/ Bases

$50.59 l

3/4.3.7.2 6elemic Monitorins Instrumentation UFsAR/ Bases

$50.59 h

3/4.3.7.3 meteorotonicet Monitorins Instrumentation UFSAR/ Bases

$50.59 4

3/4.3.7.4 Remote shutdown Instrumentation UFsAR/ Bases 150.59 p

3/4.3.7.5 Accident monitortne Instrumentation UrtAR/sases 650.59 4

3/4.3.7.6 source Ranse Monitors UFsAR/sases 650.59 i

3/4.3.7.7 Traversins in-core Probe system UFSAR/ Bases

$50.59 3/4.3.7.10 Loose Part Detection system UFsAR/ Bases

$50.59 9

(Seperate Submittel) l i-a

m 4

np J

170 Existing Title Relocation Relocation TS Document Control

)

l 3/4.3.7.11 Main condenser Offsas Treatment System UFSAR/sases (50.59 Explosive Gas Monitorins System j

3/4.3.8 Plant Systems Actuation Instrumentation UFSAR/sases 550.59 3/4.3.9 Turbine Overspeed Protection System UFSAR/sases 150.59 3/4.4.1.1 Recirculation Loops UFSAR/sases 650.59 3/4.4.1.3 Recirculation Loop Flow UFSAR/sases

$50.59 3/4.4.1.4 Idle Recirculation Loop Startup UFSAR/sases

$50.59 3/4.4.2.1 Safety / Relief Valves UFSAR/ Bases

$50.59 3/4.4.2.2 S/RVs low-Low Set Ftmetion UFSAR/sases 650.59 3/4.4.3.1 Leakage Detection System UFSAR/sases

$50.59 1

3/4.4.3.2 Operational Leakage UFSAR/sases

$50.59 i

i j

3/4.4.4 Chemistry UFSAR/sases

$50.59 j

3/4.4.5 Specific Activity UFSAR/sases 150.59 l

3/4.4.6.1 Pressure / Temperature Limits UFSAR/sases 550.59 3/4.4.7 Main Steam Line Isolation Valves UFSAR/sases

$50.59 3/4.4.8 Structural Integrity UFSAR/sases

$50.59 3/4.4.9.1 RNR Not Shutdown UFSAR/sases

$50.59 s

3/4.4.9.2 RNR Cold Shutdown UFSAR/sases 550.59 3/4.5.1 ECCS Operating UFSAR/sases

$50.59 i

3/4.5.2 ECCS - Shutdown UFSAR/sases 650.59 3/4.5.3 Stepression Pool UFSAR/ Bases 650.59

~

i i

3/4.6.1.1 Primary Contairunent Integrity UFSAR/sases 150.59 l

3/4.6.1.2 containment Leakage UFSAR/sases

$50.59 3/4.6.1.3 contalrunent Air Locks UFSAR/ Bases

$50.59 9

3/4.6.1.4 Main Steam Isolation valve Leakage control UFSAR/sases

$50.59 System 3/4.6.1.5 Feedwater Leakage control System UFSAR/sases

$50.59 3/4.6.1.8 Contalrunent Air Temperature UFSAR/sases 150.59 l

3/4.6.2.2 Drywell sypass Leakage UFSAR/ Bases 650.59 1

1 1

3/4.6.2.3 Drywelt Air Lock UFSAR/ Bases 150.59 3/4.6.2.4 Drywell Structural Integrity UFSAR/sases 150.59 1

1 3/4.6.2.6 Drywell Average Air Temperature UFSAR/sases 150.59 j

3/4.6.3.1 Suppression Pool UFSAR/sases 150.59 3/4.6.3.2 containment Spray UFSAR/sases

$50.59

l e

2 171 Existing Title Relocation Relocation 4

TS Document Control 3/4.6.3.3 Suppression Pool Cooling UFSAR/sases

$50.59 1

3/4.6.3.4 Suppression Pool Makeup UFSAR/ Bases 150.59 4

3/4.6.4 Contaltw. ant and Drywell Isolation Valves UFSAR/sases 150.59 3/4.6.5 Drywell vacuun Relief UFSAR/sases

$50.59 3/4.6.6.1 secondary Contalrunent Integrity UFSAR/sases (50.59 3/4.6.6.2 Secondary contalrunent isolation valves UFSAR/ Bases 550.59 3/4.6.6.3 Stancby Ces Treatment System UFSAR/sases

$50.59

]

3/4.6.7.1 Contaltunent Hydrogen Recombiner Systema UFSAR/sases

$50.59 i

3/4.6.7.2 contalrument and Drywell Hydrogen Ignition UFSAR/ Bases

$50.59 l

System 3/4.6.7.3 Canbustible cas Control System UFSAR/sases

$50.59 3/4.7.1.1 Service Water System UFSAR/ Bases 650.59 3/4.7.1.2 High Pressure Core Spray Service Water System UFSAR/sases

$50.59 l

3/4.7.1.3 Ultimate Heat Sink UFSAR/sases

$50.59

)

3/4.7.2 control Room Emergency Filtration System UFSAR/sases 150.59 j

3/4.7.4 Sn@bers UFSAR/sases 150.59 3/4.7.5 Sealed Source contamination UFSAR/sases 550.59 3/4.7.8 Area leaperature Monitoring UFSAR/sases

$50.59 4

3/4.7.9 Spent Fuel Storage Pool Temperature UFSAR/ Bases 650.59 3/4.7.10 Flood Protection UFSAR/ asses

$50.59 3/4.8.1.1 AC Sources Operating UFSAR/sases 650.59 3/4.8.1.2 AC Sources - Shutdown UFSAR/sases 650.59 3/4.8.2.1 DC Sources Operating UFSAR/sases 550.59 3/4.8.2.2 DC Sources - Shutdown UFSAR/ Bases

$50.59 3/4.8.3.1 Onsite Power Distribution Systems - Operating UFSAR/sases 150.59 3/4.8.3.2 onsite Power Distribution Systema Shutdown UFSAR/ Bases 650.59 1

1 3/4.8.4.1 Contalrunent Penetration Conchactor Overcurrent UFSAR/Beses

$50.59

~

Protective Devices 4

3/4.8.4.2 Motor Operated Valves Thermal Overload UFSAR/ Bases

$50.59 Protection 4

3/4.9.1 Reactor Mode Switch UFSAR/ Bases

$50.59

=

3/4.9.2 Instrumentation UFSAR/ Bases

$50.59 3/4.9.4 Decay Time UFSAR/ Bases 150.59 3/4.9.5 Communications UFSAR/sases 550.59 3/4.9.6.1 Refueling Platform UFSAR/sases

$50.59

a 172 Existing Title Relocation Relocation TS Document Control 3/4.9.6.2 Auxiliary Platform UFSAR/sases

$50.59 3/4.9.7 Crane Travet UFSAR/ Bases

$50.59 3/4.9.8 Water Level - Reactor Vesset UFSAR/ Bases

$50.59 3/4.9.9 Water Level Spent Fuel Storage and Upper UFSAR/ Bases

$50.59 Containment Fuet Storage Pools 3/4.9.10.1 Single Control Rod Removal UFSAR/ Bases

$50.59 j

3/4.9.10.2 Multiple Control Rod Removal UFSAR/ Bases 150.59 3/4.9.11.1 Reeldual Neat Removal and Coolant Circulation UFSAR/ Bases

$50.59

- Nigh Water Level 1

3/4.9.11.2 Residual Heat Removal and Coolant Circulation UFSAR/sases

$50.59 i

- Low Water level 3/4.9.12 Norizonal Fuel Transfer System UFSAR/ Bases

$50.59 3/4.11.1 Liquid Effluents UFSAR/ Bases 150.59 3/4.11.2.6 Explosive Gas Mixture UFSAR/ Bases 650.59 3/4.11.2.7 Main condenser UFSAR/ Bases

$50.59 5.2.1 Containment configuration UFSAR/sasee 150.59 5.2.2 Design Temperature and Pressure UFSAR/ Bases

$50.59 1

t 5.2.3 Secondary containment UFSAR/ Bases

$50.59 f

5.4.1 Reactor Coolant System Design leaperature and UFSAR/ Bases

$50.59 i

Pressure 5.4.7 Reactor Coolant System Volume UFSAR/ Bases

$50.59 5.5.1 Meteorological Tower Location UFSAR 550.59 5.7.1 Component cyclic or Transient Limit UFSAR 650.59 i

6.1.2 Line of Authority Directive UFSAR 550.59 6.2.2.a, Minimun Shif t Crew composition Single Unit UFSAR 550.54(a)

Table 6.2.2-Facility and SRO requirements

$50.59 1,

i j

6.2.2.g j

6.2.3 Independent Safety Engineering Grote UFSAR 550.59 6.3 Unit Staff Qualifications UFSAR 550.59 I

\\

6.4 Training UFSAR 550.59 550.55 j

1 6.5 Review and Audit UFSAR 550.54(a)

QA Plan 650.59 6.6.b Reportable Event Action UFSAR 550.59 550.73

].

6.7 Safety Limit violation UFSAR 550.72 550.73 l

650.5&

l I

L i

1 l

173 Existing Title Relocation Relocation TS Document Control 6.8.1.f Fire Protection Program Implementation UFSAR 550.48 i_

550.59 LC 2.C.(41) i I

6.8.1 3 Process Centrol Program laptementation UFSAR 550.59 6.8.1.1 Regulatory Guide 4.15 UF8AR 550.59 j

6.8.2 Review and Approvet Process UFSAR 550.54(a) 1 QA Plan 150.59 t

l 6.8.3.b In Plant Radiation Monitoring UFsAR 150.59

$20 s

l 6.8.3.e Radiological Environmental Monitoring Program UFSAR 550.59 i

6.9.1 start e Reports UFSAR 550.59 h

6.9.2 Special Reports 10CFR 550.4

~

UFSAR/Seses

$50.59 650.72 i

550.73

{

6.10 Record Retention.

UFSAR 550.54(a) i QA Plan 650.59 3

120

$71 6.11 Radiation Protection Program UFSAR 650.59 520 f

6.13 Process control Program UFSAR '

$50.59

~

$20 661 i

671 6.14 Offsite Dose Calculation Manuel UFSAR 550.59 1

l UFSAR - Updated Final Safety Analysis Report j

LC 2.C.(41) - License Condition 2.C.(41) l 4

j 2

3 1

4 5

s l

l

.