ML20045C811

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Proposed Tech Specs Section 5.0 Re Administrative Controls
ML20045C811
Person / Time
Site: 05200001
Issue date: 06/18/1993
From:
GENERAL ELECTRIC CO.
To:
Shared Package
ML20045C808 List:
References
NUDOCS 9306240407
Download: ML20045C811 (43)
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Text

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Responsibility l 5.1 5.0 ADMINISTRATIVE CONTROLS i 5.1 Responsibility 5.1.1. The [ Plant Superintendent] -shall be responsible. for overall unit operation and shall' delegate in writing the succession to this responsibility during his absence.

The [ Plant Superintendent], or his designee, in accordance with approved administrative procedures, shall approve, prior to implementation, each proposed test or experimert and proposed changes and modifications to unit systems or equipment that affect nuclear safety.

5.1.2 The [ Shift Supervisor (55)] shall be responsible for the control-  !

room command function. A management directive to this effect, 1 signed by the-[ highest level of corporate or site management]- l shall be issued annually- to all station personnel.. During any l absence of the [SS] from the control room while the unit is in i MODE 1, 2, or 3, an individual with a valid Senior Reactor '

Operator (SRO) license shall be designated to assume the control. -

room connand function. During any absence of the [SS) from'the I control room while th anit is in MODE 4 or 5, an individual lwith a valid SRO license v !.eactor Operator license shall be designated to assume tne control room command function.

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BWR/6 STS 5.0-1 Rev. O, 09/28/92 g620407930618 p) DOCK 0500000j 4-PDR f.

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, 2 Organization 5.2 5.0 ADMINISTRATIVE CONTROLS l 5.2 Organization i

5.2.1 Onsite and Offsite Oraanizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.

a. Lines of authority, responsibility, and consnunication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and

, updated, as appropriate, in organization charts, functional 1 descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the [FSAR];

b. The [ Plant Superintendent] shall be responsible for overall safe operation of the plant and shall have control over

, those onsite activities necessary for safe operation and maintenance of the plant;

c. The [a specified corporate executive position] shall have corporate responsibility for overall plant nuclear safety  ;

and shall take any measures needed to ensure acceptable '

performance of the staff in operating, maintaining, and i providing technical support to the plant to ensure nuclear safety; and

d. The individuals who train the operating staff, carry out health physics, or perfonn quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

5.2.2 Unit Staff The unit staff organization shall be as follows:

a. Each on-duty shift shall be composed of at least the minimum shift crew composition shown in Table 5.2.2-1.

i (continued)

BWR/6 STS 5.0-2 Rev. O, 09/28/92

w Organisation 5.2 5.2 Organization 5.2.2 Unit Staff (continued) _

b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, while the unit is in MODE 1, 2, or 3, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.
c. A [ Health Physics Technician] shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, l provided innediate action is taken to fill the required j position.
d. Either a licensed SRO or licensed SRO limited to fuel handling who has no concurrent responsibilities during this operation shall be present during fuel handling and shall directly supervise all-CORE ALTERATIONS.
e. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who.perfonn safety related functions (e.g., licensed SR0s, licensed R0s, health ph sicists, auxiliary operators, and key maintenance personnel .

l Adequate shift coverage shall be maintained without routine

! heavy use of overtime. The objective shall be to have operating personnel work an [8 or 12] hour day, nominal 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> week, while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed:

1. An individual should not be pemitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, excluding shift turnover time;
2. An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7 day period, all excluding shift turnover time; l

(Continued)

BWR/6 STS 5.0-3 Rev. O, 09/28/92

l Organization j 5.2 1

5.2 Organization I 4 5.2.2 Unit Staff (continued) _

! 3. A break of at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> should be allowed between i j work periods, including shift' turnover time; )

4. Except during extended shutdown periods, the use of  !

overtime should be considered on an individual basis 4 and not for the entire staff on a shift.

Any deviation from the above guidelines shall be authorized  !

in advance by the [ Plant Superintendent] or his designee, in j accordance with approved administrative procedures, or by 2

higher levels of management, in accordance with established {

procedures and with documentation of the basis for granting  !

, the deviatica. '

i 4 Controls shall be included in the procedures such that l

individual overtime shall be reviewed monthly by the [ Plant i j Superintendent] or his designee to ensure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized. _

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j The amount of overtime worked by unit staff members a

perfoming safety related functions shall be limited and

, controlled in accordance with the NRC Policy Statement on i i

working hours (Generic Letter 82-12). _

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f. The [0perations Manager or Assistant Operations Manager]

shall hold an SRO license, i

g. The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Supervisor (SS) in the areas i

of thermal hydraulics, reactor engineering, and plant j analysis with regard to the safe operation of the unit.

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BWR/6 STS 5.0-4 Rev. O, 09/28/92 4

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l Table 5.2.2-1.(page 1 of'1) '

i MinimumShiftCrewComposition(a) 4

[ Single Unit Facility]

POSITION (b) MINIMUM CREW NUMBER l UNIT IN MODE 1, 2, OR 3 UNIT IN MODE 4 OR 5 i

! SS 1 1 i

l SRO R0

$ None 1 l l l' .:

j A0(c)

STA 1 None )

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{ (a) The shift crew composition may be one less than the minimum requirements of Table 5.2.2-1 for not. more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to acconnodate unexpected j absences of on-duty shift crew members provided immediate action is. ~

taken to restore the shift crew ccaposition to within the minimum -

requirements of Table 5.2.2-1. This provision does not pennit any shift

]

j crew position to be unmanned upon shift: change due-to an oncoming shift 1 crewman being late or absent.

(b) Table Notation':

a I

[ SS - [ Shift Supervisor] with a Senior Reactor Operator license;

SRO - Individual with a Senior Reactor Operator license; i RO - Individual with a Reactor Operator license; j AO - Auxiliary Operator;
STA - Shift Technical Advisor. ,
(c) The STA position may be filled by an on-shift SS or SRO provided the

! individual meets the Consnission Policy. Statement on Engineering

, Expertise on Shift.

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Organizat.on l

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& 5'2 i Table 5.2.2-1 (page 1 of 1) j Minimum Shif t Crew Composition (a)  !

[Two Units With a Connon Control Room)

(Totals for Both Units).

i f TION (b) MINIMUM CREW NUMBER t

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ONE UNIT IN MODE 1, 2, OR 3, AND 0 EACH UNIT IN MODE UNIT IN MODE 4 0 5 EACH UNIT IN MODE

, 1, 2, OR 3 OR DEFUELED 4 OR 5 OR DEFUELED u.

l SS 1 1 1 i SR0 1 1 None RO 3 2 l AD 3 3 3 STAIC) 1 1 None f

i (a) The shift crew composition may be ne less an the minimum requirements of

] Table 5.2.2-1 for not more than 2 urs tg/ accommodate unexpected abssnee of on-duty

, shift crew members, provided f aunedia aption is taken to restore the shift crew j composition to within the minimum req cments of Table 5.2.2-1. This provision does ,

i not permit any shift crew position to unmanned upon shift change due to an oncoming j shift crewman being late or absent.

I i (b) Table Notation:

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i SS -[ShiftSupervisor)with Senior Reactor erator license for each unit whose i reactor contains fuel.

l SRO - Individual with a Sen r Reactor Operetor lic se for each unit whose reactor j contains fuel. Oth ise, provide an individua for each unit who holds a Senior Reactor Ope tor license for the unit ass ned. During CORE ALTERATIONS i on either unit a east one licensed SR0 or licens SR0 limited to fuel' )

handling, who h no other concurrent responsibiliti , must be present.

l RO - Individual wi a Reactor Operator license or a Senio eactor Operator license l

. for unit ass' ned. At least one R0 shall be assigned t each unit whose reactor

contains f I and one R0 shall be assigned as relief oper or for unit (s) in i MODE 1, 2 or 3. Individuals acting as relief operators sh I hold a license j for bot units. Otherwise, for each unit, provide a relief o retor who holds a
licens for the unit assigned.
AO - At -1 st one auxiliary operator shall be assigned to each unit se reactor e ains fuel. l STA - 5 ft Technical Advisor. ]

l i (c) The A position may be filled by an on-shift SS or SR0 provided the individ 1 meets j th Counission Policy Statement on Engineering Expertise on Shift.

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t Organization j 5.2 2 -

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! Table 5.2.2-1 (page 1 of 1) i Minimum Shift Crew Composition (a) l [Two Units With Two Control Rooms]

(Numbers for Each Unit) l POSITION (b) MINIMUM CREW NUMBER i

UNIT IN MODE

' UNIT IN N0DE 1, 2, OR 3; UNIT IN MODE 1, 2, OR 3; UNIT IN MODE OTHER UNIT 4 OR 5; OTHER i OTHER UNIT IN 4 OR 5; OTHER IN MODE UNIT IN MODE l MODE UNIT IN MODE 4 OR 5 OR 4 OR 5 OR i 1, 2, OR 3 1, 2, OR 3 DEFUELED DEFUELED i

4 SS 1(d) 1(d) 1(d) -1(d)

SRO 1 None 1 g None 1 2(e)

STA(c) 1(d) None 1 None i

i (a) The shift crew composition may be one less than the minimum requirements i

of Table 5.2.2-1 for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accomodate

unexpected absence of on-duty shift crew members provided imediate action is taken to restore the shift crew composition to within the  !
minimum requirements of Table 5.2.2-1. This provision does not pemit '

i any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

(b) Table Notation:  !

SS - [ Shift Supervisor] with a Senior Reactor Operator license; i SRO - Individual with a Senior Reactor Operator -license; j RO - Individual with a Reactor Operator license; AO - Auxiliary Operator;  !

j STA - Shift Technical Advisor.

1 (c) The STA position may be filled by an on-shift SS or SRO provided the individual meets the Commission Policy Statement on Engineering Expertise on Shift.

(d) Individual may fill the same position on the other unit if licensed for

} both.

1 (e) One of the two required individuals may fill the same position on the other unit.

BWR/6 STS 5.0-7 Rev. O, 09/28/92

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Unit Staff Qualifications 5.3 5.0 ADMINISTRATIVE CONTROLS 5.3 Unit Staff Qualifications Reviewer's Note: Minimum qualifications for members of the unit staff shall be specified by use of an overall qualification statement referencing an ANSI Standard acceptable to the NRC staff or by specifying individual position >

qualifications. Generally, the first method is preferable; however, the l l second method is adaptable to those unit staffs requiring special qualification statements because of unique organizational structures.

5.3.1 Each member of the unit staff shall meet or exceed the minimum t

qualifications of [ Regulatory Guide 1.8, Revision 2,1987, or more

! recent revisions, or ANSI Standard acceptable to the NRC staff].

l The staff not covered by [ Regulatory Guide 1.8] shall meet or l exceed the minimum qualifications of [ Regulations, Regulatory l Guides, or ANSI Standards acceptable to NRC staff]. In addition, l the Shift Technical Advisor shall meet the qualifications  !

specified by the Commission Policy Statement on Engineering 1 Expertise on Shift. '

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l Training

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5.0 ADMINISTRATIVE CONTROLS 5.4 Training 5.4.1 A retraining and replacement training program for the unit staff 4 shall be maintained under the direction of the [ position title]

and shall meet or exceed the requirements and recommendations of Section [ ] of [an ANSI Standard acceptable to the NRC staff] and 10 CFR 55, and, for appropriate designated positions, shall  ;

include familiarization with relevant industry operational '

experience.

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BWR/6 STS 5.0-9 Rev. O, 09/28/9;

Reviet:s and Audits 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Reviews and Audits Reviewer's Note: The licensee shall describe the method (s) established tc conduct independent reviews and audits. The methods may take a range of foms acceptable to the NRC. These methods may include creating an organizational unit or a standing or ad hoc committee, or assigning individuals capable of conducting these reviews and audits. When an individual performs a review function, a cross disciplinary review determination is necessary. If deemed necessary, such reviews shall be performed by the review personnel of the appropriate discipline. Individual reviewers shall not review their own work.

Regardless of the method used, the licensee shall specify the functions, organizational arrangement, responsibilities, appropriate ANSI /ANS 3.1-1981 qualifications, and reporting requirements of each functional element or unit that contributes to these processes.

Reviews and audits of activities affecting plant safety have two distinct elements. The first element is the reviews performed by plant staff personnel d

to ensure that day to day activflies are conducted in a safe manner. These reviews are described in Section 5.5.1. The second element, described in Section 5.5.2, is the [offsite] reviews and audits of unit activities and programs affecting nuclear safety tt.at are performed independent of the plant staff. The [offsite] reviews and audits should provide integration of the reviews and a dits into a cohesive program that provides senior level utility management with an assessment of facility operation and recommends actions to improve nuclear safety and plant reliability. It should include an assessment

_of the effectiveness of reviews conducted according to Section 5.5.1. _

5.5.1 Plant Reviews Reviewer's Note: The licensee shall describe provisions for plant reviews (organization, reporting, records) and the appropriate

_ ANSI /ANS Standard for personnel qualification. _

5.5.1.1 Functions The [ plant review method specified in Specification 5.5.1] shall, as a minimum, incorporate functions thet:

a. Advise the [ Plant Superintendent] on all matters related to nuclear safety; (continued)

BWR/6 STS 5.0-10 Rev. O, 09/28/92

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i Reviews and AudMs 5.5 l

{ 5.5 Reviews and Audits I 5.5.1.1 Functions (continued) f b. Recommend to the [ Plant Superintendent) approval or disapproval of items considered under Specifications 5.5.1.2.a through 5.5.1.2.e. prior to their implementation, i i

except as provided in Specification 5.7.1.3; l i

c. Determine whether each item considered under Specifications I
5.5.1.2.a through 5.5.1.2.d constitutes an unreviewed safety l question as defined in 10 CFR 50.59; and 1 l

! d. Notify the [Vice President-Nuclear Operations) of any I

safety significant disagreement between the [ review  !
organization or individual specified in Specification 5.5.1]  !

and the [ Plant Superintendent] within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. However,

the [ Plant Superintendent) shall have responsibility for )

resolution of such disagreements pursuant to -

Specification 5.1.1.

t 5.5.1.2 Responsibilities The [ plant review method specified in Specification 5.5.1]. shall 4

be used to conduct, as a minimum, reviews of the following:

l a. All proposed procedures required by Specification 5.7.1.1 j and changes thereto; i l 4

b. All proposed programs required by Specification 5.7.2 and ,

changes thereto; )

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c. All proposed changes and modifications to unit systems or equipment that affect nuclear safety;
d. All proposed tests and experiments that affect nuclear safety; and i
e. All proposed changes to these Technical Specifications (TS),

their Bases, and the Operating License.

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(Continued)

BWR/6 STS 5.0-11 Rev. O, 09/28/92

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Reviews and Audits 1 5.5 5.5 Reviews and Audits (continued)

! 5.5.2 [0ffsitel Review and Audit Reviewer's Note: The licensee shall describe the provisions for reviews and audits independent of the plant's staff (organization, i reporting, and records) and the appropriate ANSI /ANS Standards _for

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personnel qualifications. These individuals may be located onsite

' or offsite provided organizational independence from plant staff

. is maintained. The [ technical] review responsibilities, Specification 5.5.2.4, shall include several individuals located 1 onsite.

5.5.2.1 Functions '

l The [offsite review and audit provisions specified in Specification 5.5.2) shall, as a minimum, incorporate the

! following functions that: i t a. Advise the [Vice Presidant-Nuclear Operations] on all >

matters related to nuclear safety; i b. Advise the management of the audited organization, and [its i Corporate Management and Vice President-Nuclear Operations),

! of the audit results as they relate to nuclear safety; i

} c. Recommend to the management of the audited organization, and 4 its management, any corrective action to improve _' nuclear safety and plant operation; and

d. Notify the [Vice President-Nuclear Operations] of any safety significant disagreement between the [ review organization or indivioual specified in Specification 5.5.2] and the

[ organization or function being reviewed] within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.  :

5.5.2.2 [0ffsite] Review Responsibilities

! The [ review method specified in Specification 5.5.2] shall be responsible for the review of: l l .

a. The safety evaluations for changes to procedures, equipment,

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or systems, and tests or experiments completed under the

' provisions of 10 CFR 50.59, to verify that such actions do a not constitute an unreviewed safety question as defined in 10 CFR 50.59;

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i (continued) i BWR/6 STS 5.0-12 Rev. O, 09/28/92

Reviews and Audits 5.5 5.5 Reviews and Audits i

I 5.5.2.2 [0ffsite] Review Responsibilities (continued) _

b. Proposed changes to procedures, equipment,'or systems that involve an unreviewed safety question as defined in I 10 CFR 50.59; j
c. ' Proposed tests or experiments that involve an unreview:d l safety question as defined in 10 CFR 50.59 1
d. Proposed changes to TS and the Operating License .

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e. Violations of codes, regulations, orders, license  !

requirements, and internal procedures or instructions having nuclear safety significance; i

f. All Licensee Event Reports required by 10 CFR 50.73;
g. Plant staff performance;
h. Indications of unanticipated deficiencies in any aspect of i design or operation of structures, systems, or components l that could affect nuclear safety;
i. Significant accidental, unplanned, or uncontrolled radioactive releases, including corrective action to prevent recurrence;
j. Significant operating abnormalities or deviations from normal and expected performance of equipment that affect nuclear safety; and
k. The performance of the corrective action' system.

Reports or records of these reviews shall be forwarded to the

[Vice President-Nuclear Operations] within 30' days following completion of the review. -

5.5.2.3 Audit Responsibilities The audit responsibilities shall encompass:

a. The conformance of unit operation to provisions contained within the TS and applicable license conditions;
b. The training and qualifications of the unit staff; (continued)

BWR/6 STS 5.0-13 Rev. O, 09/28/92

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Reviews and Audits j 5.5 t 1 5.5 Reviews and Audits

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5.5.2.3 Audit Responsibilities (continued) _

c. The implementation of all programs required by 4

Specification 5.7.2;

, d. Actions taken to correct-deficiencies occurring in equipment, structures, systems, components, or method of operation that affect nuclear safety; and

e. Other activities and documents as requested by the [Vice
President-Nuclear Operations] .

l Reports or records of these audits shall be forwarded to the [Vice i i President-Nuclear Operations] within 30 days following completion l of the review. j 4 l 5.5.2.4 [ Technical] Review Responsibilities j I The [ technical] review responsibilities shall encompass:

j a. Plant operating characteristics, NRC issuances, industry

, advisories, Licensee Event Reports, and other sources that l j may indicate areas for improving plant safety; I

b. Plant operations, modifications, maintenance, and i surveillance to verify independently that these activities i

] are performed safely and correctly and that human errors are j reduced as much as practical; R

c. aternal and external operational experience infonnation
that may indicate areas for improving plant safety; and i d. Making detailed reconnendations through the [Vice President-Nuclear Operations] for revising procedures, equipment modifications, or other means of improving nuclear
safety and plant reliability.

l 5.5.3 Records j Written records of reviews and audits shall be maintained. As a minimum these records shall include: ,

a. Results of the activities conducted under the provisions of Section 5.5; (continued)

BWR/6 STS 5.0-14 Rev. O, 09/28/92

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i Reviews and Audits i

5.5 l

5.5 Reviews and Audits  ;

5.5.3 Records (continued) .

b. Recommendations to the. management of' the organization being audited; ,

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c. An assessment of the safety significance,of the review or l audit findings; j
d. Recommended approval. or disapproval' of items considered under Specifications 5.5.1.2.a through 5.5.1.2.e; and
e. Determination whether each item considered under Specifications 5.5.1.2.a through 5.5.1.2.d constitutes an unreviewed safety question as defined in 10 CFR 50.59.

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TS Bases Contr71 5.6 s

5.0 ADMINISTRATIVE CONTROLS 5.6 Technical Specifications (TS) Bases Control i 5.6.1 Changes to the Bases of the TS shall be made under appropriate administrative controls and reviewed according to Specification 5.5.1.

5.6.2 Licensees may make changes to Bases without prior NRC approval

, provided the changes do not involve either of the following:

a. A change in the TS incorporated in the license; or  !

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b. A change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.

5.6.3 The Bases Control Program shall contain provisions to ensure that

, the Bases are maintained consistent with the FSAR.

5.6.4 Proposed changes that meet the criteria of (a) or (b) above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71.

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BWR/6 STS 5.0-16 Rev. O, 09/28/92

i Procedures, Programs, and Manuals

. 5.7 '

5.0 ADMINISTRATIVE CONTROLS l

5.7 Procedures, Programs, and Manuals j l

J 5.7.1 Procedures l 5.7.1.1 Scope Written procedures shall be established, implemented, and

maintained covering the following activities

! a. The applicable procedures recommended in Regulatory l Guide 1.33, Revision 2, Appendix A, February 1978;  !

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i b. The emergency operating procedures required to implement the  !

requirements of NUREG-0737 and to NUREG-07 7, Supplement 1, i as stated in [ Generic Letter 82-33]; 1
c. Security plan implementation;
d. Emergency plan implementation; i
e. Quality assurance for effluent and environmental monitoring; i
f. Fire Protection Program implementation; and
g. All programs specified in Specification 5.7.2.

2 5.7.1.2 Review and Approval Each procedure of Specification 5.7.1.1, and changes thereto,

] shall be reviewed in accordance with Specification 5.5.1, approved by the [ Plant Superintendent] or his designee in accordance with 1

approved administrative procedures prior to implementation and  ;

i reviewed periodically as set forth in administrative procedures.

5.7.1.3 Temporary Changes

Temporary changes to procedures of Specification 5.7.1 may be made j provided

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a. The intent of the existing procedure is not altered;
b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator license on the unit affected; and (continued)

. BWR/6 STS 5.0-17 Rev. O, 09/28/92

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Procedures, Programs, and Manuals 5.7' 5.7 Procedures, Programs, and Manuals 5.7.1.3 Temporary Changes (continued) -

c. The change is documented ~and reviewed in accordance with' Specification 5.5.1 and approved by the.[ Plant Superintendent] 'or his designee in accordance with' approved administrative procedures within 14 days of implementation.

5.7.2 Procrams and Manuals The following programs ~shall be established, implemented, and' maintained.-

5.7.2.1 Radiation Protection Program i

Procedures for personnel radiation, protection shall be prepared 1 consistent with the requirements of 10 CFR 20 and shall be approved, maintained, and adhered to for.all operations involving a personnel radiation exposure.-

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j 5.7.2.2 Process Control. Program (PCP)'

The PCP shall contain the current formulas, sampling, analyses,

- tests, and deteminations to be made to ensure that processing and packaging of solid radioactive wastes will be accomplished to ensure compliance'with 10 CFR 20, 10 CFR 61, and 10 CFR 71;' state-regulations; burial ground requirements; and other requirements.

l governing the disposal of solid radioactive waste.

Licensee initiated changes to the PCP:

a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:-

l 1. sufficient information to' support the change (s) and l appropriate analyses or evaluations justifying the change (s),and l

l 2. a determination that the change (s) maintain the overall l conformance.of the solidified waste product to the ,

existing requirements of Federal State, or other d

applicable regulations.

i l

l' (continued)

BWR/5 STS 5.0-18. Rev. O, 09/28/92 i

L

,m--+- e '-e--

Procedures, Programs, and Manuals 5.7 i

5.7 Procedures, Programs, and Manuals 5.7.2.2 Process Control Program (PCP) (continued)

b. Shall be effective after review and acceptance by the ,

[ review method of Specification 5.5.1] and the approval of the [ Plant Superintendent].

5.7.2.3 Offsite cose Calculation Manual (0DCM) ]

a. The ODCM shall contain the methodology and parameters used )

in the calculation of offsite doses resulting from  !

radioactive gaseous and liquid effluents, in the calculation '

of gaseous and liquid effluent monitoring alam and trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program; and

b. The ODCM shall also contain the Radioactive Effluent i Controls and Radiological Environmental Monitoring programs required by Specification 5.7.2, and descriptions of the infomation that should be included in the Annual Radiological Environmental Operating, and Radioactive l Effluent Release, reports required by Specification l

[5.9.1.3] and Specification [5.9.1.4] .

Licensee initiated changes to the ODCM:

a. Shall be documented and records of reviews perfomed shall be retained. This documentation shall contain:
1. sufficient infomation to support the change (s) together with the appropriate analyses or evaluations justifying the change (s), and
2. a detemination that the change (s) maintain the levels.

of radioactive effluent control required by 10 CFR 20.106, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the

accuracy or reliability of effluent, dose, or setpoint

! calculations;

b. Shall become effective after review and acceptance by the

[ review method of Specification 5.5.1] and the approval of l the [ Plant Superintendent]; and l

(continued)

BWR/6 STS 5.0-19 Rev. O, 09/28/92 l

i Procedures, Programs, and Manuals j 5.7 5.7 Procedures, Programs, and Manuals 1

5.7.2.3 Offsite Dose Calculation Manual (0DCM) (continued) _

I

c. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of, or concurrent i with, the Radioactive Effluent Release Report for the period i of the report in which any change in the ODCM was made.

Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month

] and year) the change was implemented.

5.7.2.4 Primary Coolant Sources Outside Containment This program provides controls to minimize leakage from those

.i portions of systems outside containment that could contain highly 21 radioactive fluids during a serious transient or accident to i

levels as low racticable. The system cludeMthe' Low Af,,):.PressureCore dgi' ::: High Pressure Core oray Residual Heat re Isolation Cooling, hydrogen recombiner, yf Removal, Reactor 4 6 ',p\#o process sampling, and Standby Gas Treatment W The program shall

include the following:
,t' 4
a. Preventive maintenance and periodic visual inspection requirements; and
b. Integrated leak test requirements for each system at

, refueling cycle intervals or less.

5.7.2.5 In Plant Radiation Monitoring This program provides controls to ensure the capability- to

. accurately determine the airborne iodine concentration in vital i areas under accident conditions. This program shall include the following:

a. Training of personnel;
b. Procedures for monitoring; and
c. Provisions for maintenance of sampling and analysis equipment.

) (continued) 1 1

BWR/6 STS 5.0-20 Rev. O, 09/28/92

i

Procedures, Programs, and Manuals j 5.7 5.7 Procedures, Programs, and Manuals' l

5.7.2 Procrams and Manuals (continued) ,,,

j 5.7.2.6 Post Accident Sampling  ;

i This program provides controls that ensure the capability to  !

obtain and analyze reactor coolant, radioactive gases, and ,

i particulates in plant gaseous effluents and containment atmosphere samples under accident conditions. The program shall. include the following:

i j a. Training of personnel; l l b. Procedures for sampling an'd analysis; and 1
c. Provisions for maintenance of sampling and analysis i equipment.

5.7.2.7 Radioactive Effluent Controls Program -

i j This program confoms to 10 CFR 50.36a,for the control of

radioactive effluents and for maintaining the doses to members of

, the public from radioactive effluents:as? low as reasonably

achievable. The program shall be contained in the ODCM, shall be 4

implemented by procedures, and shall include remedial actions to be taken whenever the program limits are' exceeded. The program  ;

shall' include the following elements:

j a. Limitations-on the functional capability of radioactive.  ;

i liquid and gaseous monitoring instrumentation' including

!' surveillance tests and setpoint detemination in accordance ,

l with the methodology in the ODCM;'

i

b. Limitations on the concentrations of- radioactive material l released in liquid effluents to unrestricted ' areas, l confoming to 10 CFR'20, Appendix B. Table 'II, Column 2;_

i

! c. Monitoring, sampling, and analysis of radioactive liquid and

! gaseous effluents in accordance with 10 CFR 20.106'and with j the methodology and parameters in the ODCM; i

i d. Limitations'on the annual and quarterly doses or dose i commitment to a member of the public from radioactive i materials in liquid effluents released from each unit to j unrestricted areas, confoming to 10 CFR 50, Appendix I; 1

(continued) i BWR/6 STS 5.0-21 Rev. O, 09/28/92 i

i 1

j

\

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 1 l

5.7.2.7 Radioactive Effluent Controls Program (continued) ,,

e. Detemination of cumulative and projected dose' contributions from radioactive effluents for the' current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days; i
f. Limitations on the functional capability and use'of the liquid and gaseous effluent treatment systems to ensure that , '

appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose connitment, conforming to 10 CFR 50, Appendix I;  ;

g. Limitations on the dose rate resulting from radioactive  ;

material released in gaseous effluents to areas beyond the site boundary confoming to the dose associated with 10 CFR 20, Appendix B, Table II, Column 1;

h. Limitations on the annual. and quarterly air-doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the. site boundary,'confoming to 10 CFR 50, Appendix I; C
i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate fom with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50,' Appendix I; and.

, j. Limitations on the annual dose or dose commitment to any l member of the public due to releases of radioactivity and to

, radiation from uranium . fuel cycle sources,. confoming to l 40 CFR 190.

l 5.7.2.8 Radiological Environmental Monitoring Program l This program is for monitoring the radiation and radionuclides in l the environs of the plant. The program shall provide i representative measurements of radioactivity in the highest l potential exposure pathways and verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall be contained in the ODCM, (continued)

BWR/6 STS 5.0-22 Rev. O, 09/28/92--

l i

- . - . . - . = - _ _. . . . ... . - -.

Procedures, Programs, and Manuals

'5.7 i

j 5.7 Procedures, Programs, and Manuals  !

5.7.2.8 Radiological Environmental Monitoring Program (continued),,

shall-conform to the guidance of 10 CFR 50, Appendix'I, and'shall i include the following:

a. Monitoring, sampling, analysis,L and reporting of radiation and radionuclides in the environment in accordance with the' methodology and parameters in the ODCM;
b. A Land Use Census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitoring program are made if required by the results of this-census; and-
c. ' Participation in an Interlaboratory Comparison Program to ensure that independent checks on the' precision'and accuracy of the measurements of radioactive materials in - d environmental sample matrices are performed as part of, the-quality assurance program for environmental monitoring.-

5.7.2.9 Component Cyclic or Transient Limit This program provides controls to track the FSAR, Section [ L],..

cyclic and transient occurrences to ensure that components are maintained within the. design limits.

5.7.2.10 Pre-Stressed Concrete Containment Tendon'Surve ance Program This prog rovides controls for monit ng any tendon degradation in e-stressed concrete tainments, including y b effectiveness of 1 orrosion p ction medium, to ensure I containment structural egri .- The' program shall include baseline measurements prio initial operations. The Tendon b Surveillance Program, pectio equencies, and' acceptance criteria shall be accordance wit gulatory Guide 1.35, I i Revision 3, 19 . J-The prov' ons of SR 3.0.2 and .SR 3.0.3 are appli to the j Tend urveillance Program inspection frequencies. _ / l 5.7.2.11 Inservice Inspection Program

. A---

This program provides controls for inservice inspection of ASME CV Code Class 1, 2, and 3 components, including applicable supports.  !

/ The program shall include the following:

l (continued)

L BWR/6 STS 5.0-23 Rev. O' 09/28/92 l

<e- .-y- -

,y -

w -~ s ,.w,y,, - * - sW

Procedures, Programso and Manuals 5.7 5.7 Procedures, Programs, and Manuals 1 __

^

5.7.2.11 Inservice Inspection Program (continued) .

a. Provisions that inservice inspection of ASME Code Class 1, 2, and 3 components shall be performed in accordance with ASME Boiler and Pressure Vessel Code and Addenda,Section XI, as required by 10 CFR 50.55a; _

,,JL__

b. The provisions of SR 3.0.2 are applicable to the frequencies for performing inservice inspection activities;

< c. An inservice inspection program for piping identified in NRC

Generic Letter 88-01 in accordance with the NRC staff i

positions on schedule, methods, personnel, and sample i expansion included in Generic Letter 88-01, or in accordance

. with alternate measures approved by the NRC staff; and-

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be-4 construed to supersede the requirements of any TS. __

5.7.2.12 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports. The

- program shall include the following:

)

i a. Provisions that inservice testing of ASME Code Class 1, 2,

{ ar.d 3 pumps, valves, and snubbers shall be performed in accordance with Section XI of the ASME Boiler and Pressure I Vessel Code and applicable Addenda as required by 10 CFR 50.55a;

b. Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as follows:

(continued) 9

5.0-24 Rev. O, 09/28/92 BWR/6 STS

i . .

i Procedures, Programs, and Manuals '

5.7 1

i 5.7 Procedures, Programs, and Manuals 5.7.2.12 Inservice Testing Program (continued) ._

ASME Boiler and Pressure

Vessel Code and

! applicable Addenda l~ terminology for Required frequencies inservice testing for perfonning inservice i j activities testino activities I f Weekly- At least once per- 7 days i Monthly ~At least once per -31 days:

i Quarterly or every - 1 l 3 months At.least once-per 92 days- l

Semiannually or l every 6 months At least once per 184 days

' Every 9 months. .

.At least once per 276 days

Yearly or annually At'least once per 366-days

! Biennially or every i 2 years- At least once per 731 days ,

^ .l
c. The provisions of SR 3.0.2 are applicable.to the above l l
  • required Frequencies for performing inservice testing.

j activities;

d. The provisions of SR 3.0.3 are applicable to inservice i ~ testing activities; and' l e. Nothing in the ASME Boiler and Pressure Vessel Code shall. be j construed to supersede the requirements of any TS.

5.7.2.13 Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required i i testing of Engineered Safety Feature (ESF) filter ventilation - 'l systems at the frequencies specified in [ Regulatory Guide ],

andinaccordancewithfRegulatoryGuide1.52, Revision 2;ASME  ;

i N510-1989; and Z-G S l

?

h&

Amnt i

i l

(continued) i BWR/6 STS 5.0-25 Rev. O, 09/28/92 4

, ,- , . , , , - .e,- e,- .se < .,n,w -rw -

i i . . l Procedures, Prograns, and Manuals 5.7 l 5.7 Procedures, Programs, and Manuals 5.7.2.13 Ventilation Filter Testing Program (VFTP) (continued) ,,

j a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows i a penetration and system bypass < [0.05]% when tested in accordance with 9[ Regulatory Guide 1.52. Revision 2, and ASME "510199$ at the system flowrate specified below [2 10%]:

A G-1. - (99 :

m ESF Ventilation System Flowrate )

- ~ ~

gM SeemN4I5f4O 8

. $bn5 645 YvratWd h5My - - -  ;

b. Demonstrate for each of the ESF systems that an inplace test  !

of the charcoal adsorber shows a penetration and system 'I

, bypass < [0.5]% when tested in accordance withp[ Regulatory Guide 1.52, Revision 2, and ASME-M5&t999}%t' the system flowrate specified below [2108t]: M* 1 -I v 9  ;

i ESF Ventilation System Flowrate y '

.L

_j _ _ _

c. Demonstrate for each of the ESF systems that a laboratory i test of a sample of the charcoal adsorber, when obtained as described ingRegulatory Guide 1.52, Revision 2W shows the i methyl iodicre penetration less than the value specified 4

' below when tested in accordance with [ ASTM D3803-1989] at a temperature of s [30*C] and greater than or equal to the relative humidity specified below:

ESF Ventilation System Penetration RH

- 4-7 -

'h (continued)

BWR/6 STS 5.0-26 Rev. O, 09/28/92

~ . - . . - . .-- . . - ._ __ _ _ _ - _ -

j . .

Procedures, Programs, and Manuals 5.7 I

5.7 Procedures, Programs, and Manuals  !

5.7.2.13 Ventilation Filter Testing Program (VFTP) (continued) ,,
Reviewer's Note: Allowable penetration = [100% - methyl iodide j efficiency for charcoal credited in staff; safety evaluation]/

(safety factor).

i'
Safety factor = '5,1 for systems with swa aya6sma w a sisvu s ra

.heaters..

. ~. _ , e. . . -

e rs w l d. Demonstrate-for each of the ESF systems that the pressure

drop across the combined HEPA filters, the prefilters, and i the charcoal adsorbers is less than the value specified below when tested in accordance withfRegulatory Guide'.1.52, i Revision 2, and ASME M510-1000k.at the system flowrate

! specifiedbelow[2 10%]: Ae -l-19 91 ESF Ventilation System Delta P- Flowrate l ( ,,tiol k o w N d if d N i h h b s

! 6 W a3 o % htSph y- j

e. Demonstrate that the heaters for each of the ESF systems

, dissipate the value specified below [* 10%] when. tested in - ^--

l 7 accordance with ".= %0:0100^7r

/ -- I G-t A - t991 -

i ESF Ventilation System- Wattage y

~ ~ ~

l .

. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP j test frequencies.

l l 5.7.2.14 Explosive Gas d Storage Tank Radioactivity Monitoring Program I This program provides trols r potentially explosive gas mixtures contained in the ste Gas Holdup System],- [the quantity of radioactivity cont ' ed in storage tanks or fed into the offgas treatment s es, and the qua Rity of radioactivity

{ contained ingrotected outdoor liquid ragetanks). The j gaseous ra Koactivity quantities shall be- ined following the l (continued) j BWR/6 STS 5.0-27 Rev.- 0, 09/28/92

+ - . , -~ - _-, -- ,_-

4 4

I i

l 4

i BASES FOR DELETING SECTION 5.7.2.14 The Standard Review Plant Sections 15.7.1 and 15.7.2 have been deleted and are not analyzed in SSAR Chapter 15, Sections 15.7.1 and 15.7.2. SRP Section 15.7.1 is referenced in Branch Technical Position ETSB 11-5. The ABWR waste gas treatment system has the following attributes:

1. All of the system boundary is detonation containing.
2. No single operator error or active component failure would allow the release of ,

untreated activity. l l 3. Redundant release monitors provide isolation signals that will stop the release of l activity in excess of the design limit. )

4. The pretreatment monitors provide alarms if the process activity exceeds regulatory limits.
5. The offgas system follows the design guidance of Reg. Guide 1.143.
6. The hydrogen concentration downstream of the recombiners is monitored by redundant analyzers.

There are no liquid waste tanks in the ABWR having the attirbutes described in

, technical specification section 5.7.2.14.

4 2

Procedures, Progra2s, and Manuals 5.7 5.7 Procedures, Programs, and Manuals h.7.2.14 Explosive Gas and Storage Tank Radioactivity Monitoring Program (continued) methodology in [ Branch Technical Position (BTP ETSB 11-5,

" Post ated Radioactive Release due to Waste as System Leak or Failur "] . The liquid radwaste quantities s all be determined in accorda e with [ Standard Review Plan, Sec on 15.7.3, " Postulated Radioacti Release due to Tank Failures" .

The program s all include:

a. The limits or concentrations o' hydrogen and oxygen in the 1 [ Waste Gas H idup System] and surveillance program to ensure the li 'ts are maintai ed. Such limits shall be appropriate to he system's esign criteria (i.e., whether or not the syste is desig d to withstand a hydrogen explosion);
b. A surveillance pro m o ensure that the quantity of radioactivity contain in [each into the offgas trea t system] is gas lessstorage tank than the and fed amount that would result i a ole body exposure of n 0.5 rem to any individual in unr tricted area, in the event of [an

/ c.

uncontrolled rele se of the tanks' contents); and A surveillance rogram to e gre that the quantity of radioactivity contained in alt utdoor liquid radwaste tanks ,

that are not surrounded by lin , dikes, or walls, capable l of holding (ne tanks' contents a d that do not have tank i overflows nd surrounding area drh  !

[ Liquid adwasteTreatmentSystem]jnsconnectedtothe is less than the amount that wo ld result in concentrations less than the limits of 10 CFR / 20 Appendix B, Table II, Colu 2, at the nearest pota le wa,ter supply and the nearest s face water supply in an nrestricted area, in the event of an ncontrolled r ease of the tanks' contents.

The rovisions of SR 3.0.2 and SR 3.0.3 are appli able to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.-

(continued)

BWR/6 STS 5.0-28 Rev. O, 09/28/92

Procedures, Programs, and Manuals 5.7 4

5.7 Procedures, Programs, and Manuals 5.7.2 Programs and Manuals (continued) _

5.7.2.15 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following-l

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:
1. an API gravity or an absolute specific gravity within limits, 4
2. a flash point and kinematic viscosity within limits for ASTM 2D fuel oil, 1
3. a clear and bright appearance with proper color;
b. Other properties for ASTM 2D fuel oil are within limits  ;

within 30 days following sampling and addition to storage '

tanks; and

c. Total particulate concentration of the fuel oil is s 10 mg/l when tested every 31 days in accordance witi, ASTM D-2276, Method A-2 or A-3.

5.7.2.16 Fire Protection Program This program provides controls to ensure that appropriate fire protection measures are maintained to protect the plant from fire and to ensure the capability to achieve and maintain safe shutdown in the event of a fire is maintained.

BWR/6 STS 5.0-29 Rev. O, 09/28/92 a

i SFDP i 5.8 i

i

! 5.0 ADMINISTRATIVE CONTROLS

5.8 Safety Function Determination Program (SFDP) '

4 5.8.1 This program ensures loss of safety function is detected and i

appropriate actions taken. Upon failure to meet two or more LCOs i at the same time, an evaluation shall be made to determine if loss l of safety function exists. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as.a result'of the support system inoperability and

i corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of' l LCO 3.0.6.

i j 5.8.2 The SFDP shall contain the following:

4 j a. Provisions for cross division checks to ensure a loss of'the-capability to perform the safety function assumed in-the j accident analysis does not go undetected;

, b. Provisions for ensuring the plant is maintained in a ' safe

, condition if a loss of function).ondition exists; J

c. Provisions to ensure that an inoperesle supported system's

. Completion Time is not inappropriately extended as a result j of multiple support system inoperabilities; and

d. Other appropriate limitations and remedial or compensatory
actions.

1 5.8.3 A loss of safety function exists when, assuming'no concurrent

single failure, a safety' function assumed in the accident-analysis i cannot be performed. For the purpose of this program, a loss of 3 safety function may exist when a support system is inoperable,

! and:

[ .

i

! a.. A required system redundant to system (s)- supported by the j inoperable support system is also inoperable (Case A); or  !

b. A reqdited. system redundant to system (s) in turn supported by the inoperable supported system is also inoperable-

! (Case B); or 1

I

[ (Continued).

i l BWR/6 STS 5.0-30 Rev. O, 09/28/92-l y4 e- e-ie - ,-. .--- y-n ,.w.-.r . - - - , w-y-, c p - ,-wr-a

SFDP 5.8 5.8 SFDP i

5.8.3 (continued)

c. A required system redundant to support system (s) for the i supported systems (a) and (b) above is also inoperable  !

(Case C). l 1

Generic Example: '

Division A Division B System i System i --Case C l j 4 4 System ii .-(Support System System ii 4 Inoperable) 4 System iii System iii +-Case A 4 4 System iv System iv .-Case B 5.0.4 The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

e BWR/6 STS 5.0-31 Rev. O, 09/28/92

, Reporting Requirements i 5.9 5.0 ADMINISTRATIVE CONTROLS

, 5.9 Reporting Requirements i

5.9.1 Routine Reports The following reports shall be submitted in accordance with 10 CFR 50.4.

1 1 5.9.1.1 Startup Report A summary report of plant startup and power escalation testing shall be submitted following:

j e

a. Receipt of an Operating License;
b. Amendment to the license involving a planned increase in
power level;
c. Installation of fuel that has a different design or has been manufactured by a different fuel supplier; and
d. Modifications that may have significantly altered the i nuclear, thermal, or hydraulic performance of the unit.

The initial Startup Report shall address each of the startup tests' i identified in FSAR, Chapter [14], and shall include a description l l of the measured values of the operating conditions or  ;

i characteristics obtained during the test program and a comparison )

of these values with design predictions and specifications. Any l corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific-details required in license conditions based on other commitments shall be included in this report. Subsequent Startup Reports
shall address startup tests that are necessary to demonstrate the acceptability of changes and modifications.

I Startup Reports shall be submitted within 90 days following

! completion of the Startup Test Program; 90 days following-s resumption or commencement of commercial power operation; or 3 9 months following initial criticality, whichever is earliest. If 3 the Startup Report does not cover all three _ events (i.e., initial criticality, completion of Startup Test Program, and resumption or commencement of commercial operation), supplementary reports shall be submitted at least every 3 mont3s until all three events have been completed.

(continued)

BWR/6 STS 5.0-32 Rev. O, 09/28/92

Reporting kequirements 5.9 5.9 Reporting Requirements 5.9.1 Routine Reports (continued) 5.9.1.2 Annual Reports

.......................--..--..N0TE.-.---- ---...--......-....--..

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station.

Annual Reports covering the activities of the unit as described below for the previous calendar year shall be submitted by March 31 of each year. [The initial report shall be submitted by March 31 of the year following initial criticality.]

Reports required on an annual basis include:

a. Occupational Radiation Exposure Report A tabulation on an annual basis of the number of station.

utility, and other personnel (including contractors) receiving exposures > 100 mrem /yr and their associated man rem exposure according to work and job functions (e.g.,

reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance [ describe maintenance], waste processing, and refueling). This tabulation supplements the requirements of 10 CFR 20.407.

The dose assignments to various duty functions may be

, estimated based on pocket dosimeter, themoluminescent dosimeter (TLD), or film badge measurements. Small exposures totalling < 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources should be assigned to specific major work functions; and

[b. Any other unit unique reports required on an annual basis.]

(continued)

BWR/6 STS 5.0 33 Rev. O, 09/28/92

- ._ _ ~ - .- . - - . . - - - . . ... -- . --

]

! Reporting Requirements l 5.9 '

{ 5.9 Reporting Requirements l

5.9.1 Routine Reports (continued). .

5.9.1.3 Annual Radiological Environmental Operating Report l

~

..-----.------------NOTE.-------------.----------------

! A single submittal may.be made for a multiple unit station. The i

submittal should combine sections common to all units at the j station.

j --.------ .-.--.------.----..--..-------..--..------------.-.-----

The Annual Radiological . Environmental Operating. Report covering the operation of the unit during-the previous calendar year shall be submitted by May 15 of each year. The report shall include-suinnaries, interpretations, and analyses of trends of the results l of the Radiological Environmental Monitoring Program for the-j reporting period.- The material provided shall' be consistent with

! the objectives outlined in.the Offsite Dose Calculation Manual (ODCM), and in:10 CFR 50, Appendix I, Sections IV.B.2, IV.8.3,

{ and IV.C.

The Annual' Radiological Environmental Operatin Report shall include the results of analyses of all radiolo ical environmental l

! samples and of all environmental radiation measurements-taken  !

during the period pursuant- to the locations.specified. in the table i and figures in the ODCM, as well as summarized and tabulated  ;

results of these analyses and measurements [in the fomat of the i

table in the Radiological Assessment Branch Technical Position, l Revision 1,' November 1979]. [The report shall identify the TLD

~

results that represent collocated dosimeters-in relation to the --
NRC TLD program and the exposure period associated with each  ;

result.] In the event that some individual results are not

available-for inclusion with the report, the report shall be-submitted noting and explaining the reasons for the missing '
results. The missing data shall be submitted in a supplementary
report as soon as possible.

If (continued)

BWR/6 STS 5.0-34 Rev. O,09/28/92

I e

s e .

! Reporting Requirements 5.9 2 l 5.9 Reporting Requirements  ;

1 5.9.1 Routine ReDorts (continued) .,

5.9.1.4 Radioactive Effluent Release Report i j -.-------.---------------------NOTE---------------------------.---

d A single submittal may_be made for a multiple unit station. -The i submittal should combine sections comon to all units.'at the i station; however, for units with separate radwaste systems, the I submittal shall specify the releases of radioactive material from i each unit.

i The Radioactive Effluent Release Report covering the operation of l the unit shall be submitted in accordance with 10 CFR 50.36a. ~The l report shall include a sumary of the quantities of radioactive 9 liquid and gaseous effluents and solid waste released from the 1

unit. The material provided shall be consistent.with the objectives outlined in.the ODCM and Process Control Program and in- 1 1

conformance with 10 CFR 50.36a and_10 CFR 50, Appendix.I,Section IV.B.1.

5.9.1.5 Monthly Operating Reports 4

Routinereportsof'operatingstatisticsandshutdownexperience[,

includin valves ]g documentation of all challenges to the safety /reliefshall be sub 15th of each month following the calendar month covered by the report.

5.9.1.6 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

The individual specifications that address core operating limits must be referenced here.

b. The analytical methods used to detemine'the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described-in the following documents:

.(continued)

BWR/6 STS 5.0-35 Rev. 0, 09/28/92

j . o e 4

i Reporting Requirements s

5.9 j 5.9 Reporting Requirements j  !

l 5.9.1.6 CORE OPERATING LIMITS REPORT (COLR) '

(continued) ,,

l

! Identify the Topical Report (s) by number, title, date, and i NRC staff approval document, or identify the staff Safety Evaluation Report for a plant specific methodology by NRC b

! letter and date.

1 l c.- The core operating limits shall- be detemined such that all e

applicable limits (e.g., fuel themal mechanical limits,

, core themal hydraulic limits, Emergency Core Cooling j Systems (ECCS) limits, nuclear limits such.as SDM, transient i j analysis linits, and accident analysis limits) of the safety- l l analysis are met.-  !

. 1

-d. The COLR,_ including any midcycle__ revisions or supplements, j shall be provided upon issuance for each reload cycle.to the i NRC.
5.9.1.7 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS l REPORT'(PTLR)

! The RCS pressure and temperature limits, including heatup and i cooldown rates, criticality, and hydrostatic and leak test limits,

! shall be established and documented in-the PTLR. [Theindividual i Specifications that address the reactor vessel pressure and~

temperature limits-and the heatup and cooldown rates may be-

)-

j referenced.] The analytical methods used to detemine the pressure and temperature limits including the heatup and cooldown rates shall be those previously reviewed and approved by the NRC in [ Topical Report (s), number, title, date, and NRC staff approval document, or staff safety evaluation report for a plant specific

methodology by NRC letter and date] . The reactor vessel pressure
and temperature limits. including those for heatup and cooldown rates, shall be detemined so that all applicable. limits- (e.g.,

heatup limits, cooldown limits, and inservice leak and hydrostatic testing limits) of_ the analysis are met. The PTLR, including

, revisions or supplements thereto, shall be provided upon issuance

for each reactor vessel fluency period.

(continued)

BWR/6 STS 5.0-36 Rev. O, 09/28/92 t'

b

- . . -. -- . - . -- . . _ . . . . - - a.

Reporting Requirements 5.9  ;

l 5.9 Reporting Requirements (continued) i 5.9.2 Special Reports i

Special Reports may be required covering inspection, test, and

maintenance activities. These special reports are determined on an individual basis for each unit, and their preparation and l submittal are designated in the Technical Specifications.

Special Reports shall be submitted in accordance with 10 CFR 50.4 within the time period specified for each report. l The following Special Reports shall be submitted:

a. In the event an ECCS is actuated and injects water into the RCS in MODE 1, 2, or 3, a Special Report shall be prepared and submitted within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
b. If an individual emergency diesel generator (EDG) experiences four or.more valid failures in the last 25 demands, these failures and any nonvalid failures experienced by that EDG in that time period shall be reported within 30 days. Reports on EDG failures shall include the information recommended in Regulatory Guide 1.9, Revision 3, Regulatory Position C.5, or existing Regulatory

_ Guide 1.108 reporting requirement. _

c.

When LCO 3.3.a Sp[3.1], " Post Accident Monitoring (PAM)ecial Report is Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

(continued)

BWR/6 STS 5.0-37 Rev. O, 09/28/92

a.

. s Reporting Requirements 5.9 5.9 Reporting Requirements
5.9.2 Special Reports (continued)- ._
d. Any ab i degradation of the c ainment structure q detected du ' the tests re ' ed by the Pre-Stressed Concrete Contain en urveillance Program shall be.

reported to the NRC 0 days. The report shall include a descri on of the n condition, the condition of the coner (especially at ten chorages),the

, inspect procedures, the tolerances on ing, and the

, cor ive action taken.

i i

i l I .

e d

A 9

BWR/6 STS 5.0-38 Rev. O, 09/28/92

. -. -. ..- -. .-.- - -- . - - ~ . . . - . . _ . . . . - . . . _ . . - .

s o. >

Record Retention i 5.10 l i

5.0 ADMINISTRATIVE CONTROLS 5.10 Record Retention 5.10.1 The following records shall be retained for at least 3 years:

a. All License Event Reports required by 10 CFR 50.73;
b. Records of changes made to the procedures required by Specification 5.7.1.1; and
c. Records of radioactive shipments.

5.10.2 The following records shall be retained -for at least 5 years: )

a. Records and logs of unit operation covering time intervals at each power level;
b. Records and logs of principal maintenance activities-inspections, repair, and. replacement of. principal items. of equipment related to nuclear safety;
c. Records of surveillance activities, inspections, and.

calibrations required by the Technical Specifications (TS) [and.the Fire Protection Program];

d. Records of sealed source and fission detector leak tests and results; and e.- Records of annual physical inventory of all.' sealed source material of record.

5.10.3 The following records shall be retained for the duration of the  !

unit Operating License-  !

a. Records and drawing changes reflecting unit design modifications made to systems and equipment described in the i FSAR; l
b. Records of new and irradiated fuel inventory, fuel.

transfers, and assembly burnup histories;

c. Records of radiation exposure for all individuals. entering radiation control areas;

-(continued)

BWR/6 STS 5.0-39 Rev. O, 09/28/92-

q 09 4 Record Retention 5.10 5.10 Record Retention l 5.10.3 (continued) _  ;

d. Records of gaseous' and liquid radioactive material released j to.the environs; ,
e. Records of transient or. operational cycles for those unit )

components identified in [FSAR, Section X], 1

f. Records of reactor tests and experiments; ')

9 Records of training and qualification for members of the l unit staff;

h. . Records of inservice inspections perfomed pursuant to the j TS; '
i. Records of quality assurance activities required by the Operational Quality Assurance (QA) Manual [not listed in Specification.5.10.1 and which are classified'as permanent records by applicable regulations, codes, and standards];
j. Records of reviews perfomed for changes ma'd e to procedures, equipment, or reviews of tests and experiments pursuant- to 10 CFR 50.59;
k. Records of the reviews and audits required by Specification 5.5.1 and Specification 5.5.2;
1. Records of the service lives of all hydraulic and' mechanical snubbers relocated to req)uired

, including by the

[ document wherethe_

date at which snubber service requirements life commences, and associated installation and maintenance records;

m. Records of analyses required by the Radiolog'ical Environmental Monitoring Program that would pemit-evaluation of'the accuracy of the analysis at a later date (theserecordsshouldincludeprocedureseffectiveat specified times and QA' records showing that these procedures werefollowed);

(W

n. Records of reviews perfomed for changes made to the Offsite p 'A Dose Calculation Manual and the Process Control Program;-and hD -

-~ _C

o. Records of pre-stressed concrete containment tendon surveillances.

W M

BWR/6 STS 5.0-40 Rev. O,'09/28/92

[High Radiation Area'

[5.11l 5.0 ADMINISTRATIVE CONTROLS

[5.11 High Radiation Area]

5.11.1 Pursuant to 10 CFR 20, paragraph 20.203(c)(5), in lieu of the l requirements of 10 CFR 20.203(c), each high radiation area, as  ;

defined in 10 CFR 20, in which the intensity of radiation is

> 100 mrem /hr but < 1000 mrem /hr, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g., [ Health Physics Technicians]) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performaace of their assigned duties in high radiation areas with exposurt: rates s 1000 mrem /hr, i provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.

Any individual or group of individuals permitted to enter such i areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device that continuously indicates the radiation dose rate in the area,
b. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alams when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are l aware of them.
c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the i activities within the area and shall perfom periodic radiation surveillance at the frequency specified by the

[ Radiation Protection Manager] in the RWP.

5.11.2 In addition to the requirements of Specification 5.11.1, areas with radiation levels a 1000 mrem /hr shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Foreman on duty or health physics supervision. Doors shall remain locked except during periods of access by personnel _

(continued)

BWR/6 STS 5.0-41 Rev. O, 09/28/92

i

[High Radiation Area'

[5.11[

. [5.11 High Radiation Area]

l 5.11.2 (continued)

I under an approved RWP that shall specify the dose rate leveIs in' the innediate work areas and the maximum allowable stay times for individuals in those areas. In lieu of.the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

j 5.11.3 For individual high radiation areas with radiation levels of i > 1000 mrem /hr, accessible to personnel, that are located within

large areas such as reactor containment, where no enclosure exists
for purposes of locking, or that cannot be continuously guarded, and where no enclosure can be reasonably constructed around the

~

individual area, that individual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a warning device.

1 BWR/6 STS 5.0-42 Rev. O, 09/28/92

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