ML20214K328

Summary of 860527-30 Meetings W/Util,Southern Co Svcs, Bechtel & Westinghouse at Site to Discuss Tech Specs. Supporting Documentation Encl

ML20214K328
Person / Time
Site:	Vogtle
Issue date:	08/11/1986
From:	Mark Miller Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8608210042
Download: ML20214K328 (119)

Text

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• UNITED STATES " .

n NUCLEAR REGULATORY COMMISSION

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5 , WASHINGTON, D. C. 20658 i

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, l'1 AUG 198S 1 Docket Nos.: 50-424 -

I and 50-425 '

APPLICANT: eorgia Power Company FACILITY: Vogtle, Units 1 and 2

SUMMARY

OF TECHNICAL SPECIFICATION MEETING HELD ON i MAY 27-30, 1986  :

v The staff met with the applicant and its representatives at the Vogtle site to discuss the Vogtle Technical Specifications (TS). Participants are listed in Enclosure 1. The first day was spent familiarizing the staff's TS reviewers with the various components and equipment at Vogtle. The remainder of the week .

was spent discussing the applicant's proposed Vogtle-specific changes to the WestinghouseStandardTechnicalSpecifications(STS).

Enclosure 2 contains the appifcant's meeting notes transmitted by letter dated i June 20, 1986. The staff incorporates these notes into the meeting sumary with the following clarifications and additions. The applicant's attachments as .

called out in the notes are not part of Enclosure 2  ;

Section 6.0 [

. . Enclosure 3 contains a sample TS Section 6.5.3 which the staff provided to the i applicant. Enclosure 4 contains the marked up version of Section 6.0 reflecting additional staff revisions.

Section 6.2.2.b b.O i The applicant wil retain the wording of the STS.

• Section 6.2.2.c f The staff indicated that it would be necessary to establish the context in which a health physicist (HP) would be used in order to decide if individuals other i than an HP could be used. -

Section 6.2.2.d j The applicant will retain the wording of the STS.  ;

Section 6.5.1 The proposed version of 6.5.1.6 provided to the applicant is contained in '

Enclosure 5. The staff will check the Plant Review Board (PRB) minimum qual- ,

iftcation levels. The applicant needs to indicate that PRB composition will not change monthly. 1 I

5 0609210042 060011 4 1 PDR ADOCK 0000

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2 Section 6.10.3.1 -

The staff will clarify whether this section refers to all activitihs or only j safety-related ones.

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Table 3.6-1 $

I Enclosure 6 contains a copy of a May 20, 1986, letter regarding Seabrook which >

was provided to the applicent at the meeting.

Section 4.1.2.3 '

The staff needs to look at the applicant's PRA-based boron dilution report to determine if the proposed wording of this section is acceptable. .

Section 3/4.1.3. Table 3.1-1 (Movable Control Assemblies)

The applicant provided the staff with a copy of a December 21, 1984, letter 1 from Westinghouse which supports the proposed wording of this section. The letter is in Enclosure 7.

Sections 3/4.3.3.8, 3/4.7.11, and 3/4.7.12

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Fire Suppression and Fire-Rated Assemblies) (Fire Detection Instrumentation, i i

o The applicant and staff agreed to defer discussion on this item until after the staff reviews the applicant's sending fire protection submittal at which point these sections could possibly )e deleted. t

[

Section 3/4.6.1.7 (Containment Ventilation System) O The staff provided a copy of the Palo Verde T{(Enclosure 8) in this area to the applicant. '

r Section 3/4.7.5

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Enclosure 9 provides an illustrative table for the basis of the appitcant's proposed revision to this section.

Sections 3/47.6 and 3/4.7.8 (Control Room Emergency Air Filtration and ECCS PumpRoomExhaustAirCleanupSystems)

The applicant provided a copy (Enclosure 10) of a Farley TS amendment related to t the control room emergency air filtration system and ECCS pump room exhaust air i, cleanup system. '

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11 AUG 175 Section 3/4.8.4.1 (Containment Penetration Conductor Overcurrent Protective Devices) .

A TS is required on the feeder breaker as stated in the SER, p. 8-14 Section 3/4.8.4.2 (Motor Operated Valves Thermal Overload Protection Bypass Devices)

The applicant proposes to delete 15 Table 3.8-2 as was done for Seabrook. The applicant would need to include this information in the Vogtle FSAR.

Sections 3/4.8.1.1 and 3/4.8.1.2 (ACSources)

The staff provided the applicant.with a copy of the Perry TS in this area (Enclosure 11). On p. 3/4 8-6, the staff was to clarify what test is required regarding lockout features. On p. 3/4 8-7, the staff will clarify the purpose of the test in item f. In Section 4.8.1.1.3, the staff needs to check the 30 day reporting requirement on failures.

Section 3/4.9.6 Enclosure 12 contains Vogtle-specific markups of the refueling machine TS.

Ilowever, the applicant indicated that this TS may undergo further revision.

Section 4.4.6.2.2 The staff pointed out that the wording of this TS is not consistent with the response to FSAR Q210.48 and SER p. 3-48. The applicant needs to address this.

Enclosure 13 contains the applicant's comparison of TS requirements with the SER.

\b\

t Melanie A. Miller, Project Manager

! PWR Project Directorate #4 Division of PWR Licensing-A l

Enclosures:

As stated I O PW ' WR-A PWR#4/DPWR-A

! Mit br rad BJYoungblood 08/ /86 08/t\ /86 l

MEETING

SUMMARY

DISTRIBUTION

~b d ekat Fii D~~

NRC Participants (RC PDR M. Miller L PDR S. Brown NSIC J. Rogge PRC System P. Chan PWRf4 Reading File MMiller M. Duncan OGC J. Partlow E. Jordan B. Grimes ACRS(10)

OTHERS i

bec: Licensee & Service List i

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_ Enclosure 1 l

l Participants l-NRC Southern Company Services-IC" hiller J. stringfellow S. Brown B. George J. Rogge R. Amstrong P. Chan Geor la Power Company Westinghouse

. orian K. Daschke i G. Bockhold*

J. Swartzwelder Bechtel R. Porter 5. Mahler G. Lee R. Schilling

• J. Sutphin R. Hand I. Kochery*

P. Hermann D. Hudson

• E. Cobb*

A. Desrosiers*

W. Chenault*

T. Wendt

• 0nly attended portion of meeting l

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' Enclosure 2 ATTACIDerf 1 M ETIM NOTES YBGP UNIT 1 M AFT TECWICAL SPECIFICATIONS MC SITE VISIT M&Y 27-30, 1946

• s, Section 6.0:

The reviewer distributed copies of Section 6.0 which had been marked up to reflect additional revisions accepted by the staff since the first draft was issued in typed form on May 1,1986. This hand-out served as the basis for the

• discussion of Section 6.0. Attachment 2 contains a mark-up of Section 6.0 which reflects those revisions which were agreed to at the meeting. The following itees are comments concerning specific areas of Section 6.0 or items which require additional followup.

6.2.2.ca The reviewer will give our proposed revision further consideration.

6.2.2.es

- We should be able to delete this ites when the staff has a chance to review

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our Fire Hazards Analysis (Appendix 9A to the FSAR).

i Footnote

• to 6.2.2.c l ,

If the staff accepts our proposed revision to 6.2.2.c and 6.2.2.e is deleted,

our proposal to revise the 2-hour limitation to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> any be unnecessary.

I GPC will check to see if the 4-hour limitation is reflected in the Fire Protection Program as described in the FSAR.

6.2.2.f:

The reviewer will give our proposed revision to the last two sentences of the last paragraph further consideration.

Table 6.2-1: .

The reviewer stated that since we are in the process of licensing the first unit of a two-unit plant, the Technical Specifications will have to be written so as to address only the one unit. (However, he did agree to change the word

" unit" to " plant" in certain areas of Section 6.0). Therefore, Table 6.2-1 will have to address only a single unit. GPC will revise this table to address a single unit and the defueled mode which is ignored in the Standard.

6.2.3:

GPC agreed to drop the request to delete the requirement to review LERs.

However, the NRC will provide clarification as to whether or not this includes LERa from other utilities.

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MIETING NOTES

!. YBGP, UNIT 1 ERAFT TECENICAL SPECIFICATIONS NRC'dITE VISIT M&Y 27-30. 1986 Page 2

-6.3:

• s s.

This section was deleted from the Seabrook Technical Specifications.

Consequently, once the area of Plant Staff Qualifications is resolved in the j realm of the FSAR, we should be able to delete this section from our Technical 3 Specifications.

6.4 i Once this area is resolved in the realm of the FSAR, the Technical 4

Specifications will be revised as necessary to be consistent with the FSAR.

6.5.1:

GPC pointed out that the Technical Specifications concerning the Plant Review Board were. based on a position negotiated with the former reviewer of Chapter 13 of the FSAR. The NRC agreed to go back to the former reviewer and discuss

• our proposed revisions with him. The NRC also handed out a proposed 6.5.1.6 for GPC's consideration. However, review of what was proposed by the NRC will be deferred until such time as our version of 6.5.1 is discussed with the former reviewer of Chapter 13 of the FSAR.

t 6.5.2.2 GPC pointed out that the Technical Specifications which govern the Safety i

Review Board for VEGP should be consistent with those for Hatch, since the

same SRB is used for both facilities. In light of this, our suggested

] revisions are intended to bring about this consistency. The reviewer will i discuss this further with the NRC Project Manager for Hatch.

i 6.8.2, 6.8.3.cs l

! Our proposed revisions to these items will be addressed with the resolution of j 6.5.1.

i 6.8.4.c.3:

GPC pointed out that this proposed revision is intended to be plant-specific.

l Charlie Willis, the REIS reviewer for the NRC, will consider this ites.

6.9.1.3: i l The concern here has to do with the 3rd paragraph of page 6-18. The reviewer j agreed to delete the words "at least two" with regard to legible maps but the footnote

• was retained which has the effect of requiring two asps. The more

! distant stations may require more than one map for legible coverage, whereas the footnote specifies that the second map (singular) sust cover the sore distant stations. The intent of our revision was to allow flexibility in providing maps.

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MBTING NOTES VEGP WIT 1 MAFT T3GNICAL SPECIFICATIONS RC SITE VISIT M&T 27-30, 1986 Page 3 6.10.3.m:

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• s, GPC has proposed that records of secondary water sampling and water quality be

! retained for 5 years rather than for the duration of the plant Operating License on tha basis that five-year intervals should be of sufficient length to allow identification of any negative trends. The reviewer will reconsider this position.

6.10.3.it

! The reviewer will clarify the intent of this requirement. GPC feels that the FSAR is the governing document rather than the Quality Assurance Manual.

6.12.1 and 6.12.2:

! The reviewer will have to discuss the proposed revisions with Charlie Willis,

the RETS reviewer. GPC pointed out that the revisions to 6.12.2 were -

j plant-specific in nature.

- 6.13.1 and 6.14.1:

GPC pointed out that these iteiins should be deleted on the basis that the PCP i and the ODCM have already been submitted for review and approval which will take place prior to issuance of these Technical Specifications. The reviewer

, will give this argument further consideration.

6.13.2.b, 6.14.2.b 6.15.1.a.8 and 4.15.1.b:

f These itees will be resolved with the resolution of our proposed revisions to i 6.5.1. -

6.16:

GPC agreed to drop this proposed ites.

! Definittoa 1.7, 4.6.1.-l.a and Tabla 3.6-1 l In our submittal, we had proposed that Table 3.6-1 consist of a listing of i only automatic containment isolation valves. The reviewer stated that the

staff position requires that all containment isolation valves (including

! manual valves and dual function valves) be listed. However, Seabrook has recently been successful in deleting this table from their Technical 1 Specifications by relying on a listing in the FSAR. The staff would be

willing to consider this for VEGP if we ask for it and provide the necessary justification. We would have to make sure that the FSAR is clear with regard to what we take credit for as containment isolation valves.

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O MEETING NOTES TBGP UNIT 1 DEAFT TBCENICAL SPECIFICATIONS IRC SITE VISIT MAY 27-30. 1986 Page 4 -

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4.6.1.1.a. Footnote 8 s s.

After the site tour, the reviewer stated t' hat he had a better understanding

! for our requested exception concerning the blind flange on the fuel transfer

! canal. He will take our arguments back to the branch reviewer for further consideration. l

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3.4.3: '

CPC will consider the wording used in the Perry Technical Specifications for the exception to the provisions of Specification 3.0.4. See Attachment 3 for a copy of the Perry specification.

Controlled Leakage 1

The reviewer stated that our proposed revisions concerning the limits on i

controlled leakage have also been requested for Seabrook and are currently -

under review on that docket. He suggested that we wait for the results of the l Seabrook review rather than initiating a separate review on our own docket.

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Definition of it i

l CPC will consider wording similar to that used in the Perry Technical i

Specifications for this definition. See Attachment 3 for a copy of the Perry definition.

Boron Dilution Accident Analysis Discussion of this itse was deferred until Westinghouse completes the reanalysis for this event. The reviewer requested additional information

! regarding the expected frequency of a baron dilution event simultaneous with a i loss-of-offsite power. This is related to our proposed deletion of the requirement for emergency power to the charging pumpe for reactivity control.

The intent of this proposal is to allow use of the positive displacement pump for emergency boration. The reviewer will give our proposal further

, consideration.

Cold Overpressurisation Analysis

! Concerning the Action Statements for 3.8.1.2, 3.8.2.2, and 3.8.3.2, the

! reviewer will have to give our proposed revisions to the requirement for

) venting the RCS further consideration. See Attachment 2 for those revisions j accepted by the reviewer.

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MEETING NOTES

, YEGP UNIT 1 IIRAFT TECENICAL SPECIFICATIONS NRC SITE YISIT MAY 27-30, 1986 Page 5 Movable Control Assemblies:

• s s, GPC agreed to the Diablo Canyon version of this specification provided that Table 3.1-1 is reworded to reflect the titles of the accident analyses as they appear in Chapter 15 of our PSAR.

The reviewer will give our proposed bases for this specification further consideration. ,

RCS Flow Rate and Nuclear Enthalpy Eise - Hot Channel Factor GPC will consider the approach adopted by Seabrook for Specifications 3/4.2.3, 3/4.2.4, and 3/4.2.5.

Turbina Overspeed Protection:

CPC will expedite the internal review of this program so that a submittal can

• be made to the NRC staff for their review.

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RCS Specific Activity sad Secondary Coolant Specific Activity:

The reviewer accepted a half-life cutoff of 14 minutes as opposed to 10 minutes.

Structural Integrity:

Our submittal for proposed surveillance 4.4.10.3 contains the words "to the extent practical." The reviewer suggested that we check our ISI submittal for these same words. -

Containment Structural Integrity:

We need to provide an appropriate tensile strength for the tendons. See Attachment 2 for those revisions accepted by the reviewer.

Contaissent Yeatilation Systems CPC will consider the wording used in the Palo Verde Technical Specifications for this specification. See Attachment 3 for a copy.

Main Steam Isolation Yalvest l The reviewer re<;uested that we submit the PRA which we performed in support of the allowed outage times which we requested in our submittal. He also l requested information concerning the margin which exists in our analysis.

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IEITING NOTES

, YEGP UNIT 1 DRAFT TECENICAL SPECIFICATIONS  !

NRC SITI TISIT MAY 27-30. 1986 Page 6 j Contaiument Penetration Conductor Overcurreng Protective Devices Table 3.8-1 was deleted from the Seabrook Technical Specifications and placed J in their FSAR. GPC will consider making a similar request for the VEGP Unit 1 Technical Specifications. With regard to our proposed revisions to the Action Statements, the reviewer will have to give our requests further consideration.

Notor-operated Yalves Thermal Overload Protection Bypass Devices See Attachment 2 for revisions accepted by the reviewer at the meeting. ,

A.C. Sources

! The reviewer will have to give our proposed deletion of 4.8.1.1.1.b further consideration. With regard to.4.8.1.1.2.c, whether or not the surveillance is monthly or quarterly will depend on the results of our groundwater monitoring program. The reviewer will also clarify the requirements of 4.8.1.1.2 3 13.

l What is the intent of this surveillance requirement?

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The reviewer requested that we review the Perry specifications concerning emergency A.C. sources and compare these specifications to those that we proposed in our submittal. Due to the significant differences between what we requested and what has been recently granted for plants relying on TDI diesels, the reviewer requested that we have a separate meeting with the Power

!- Systems Branch reviewer to discues our requests.

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, Water Imvel

.1 See Attachment 2, for those revisions accepted by the reviewer at the meeting.

Definition 1.29, Figure 5.1-1:

See Attachment 2 for the wording accepted by the reviewer at the meeting. We agreed to provide a more legible map for Figure 5.1-1.

3/4.9.6:

We discussed a specification based on the one granted for Millstone 3 and the reviewer agreed in principal. However, Westinghouse, Bechtel and GPC will have to work together to ensure that our proposed specification is consistent with our heavy loads analysis as it appears in the FSAR and that the load limits specified are accurate for VEGP.

4.0.2:'

The reviewer will have to give our proposed revision further consideration.

GPC will consider sodifying their request to include the standard wording and to list the intervals separately in a table.

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MEETING NOTES VEGF UNIT 1 NEAFT TECENICAL SPECIFICATIONS M C SITE VISIT M&Y 27-30. 1986 4

Page 7

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, 4.0.3: ,

The reviewer will have to give our suggested revision further consideration.

3.1.2.1.b and 3.1.2.1.at hose LCOs need to be reworded to clarify the flowpath requirements.

f 3.1.3.2, Proposed Action C:

, We explained to the reviewer that in order to verify operability following l repair, we need an exception to the provisions of Specification 3.0.4. He will give our request further consideration.

I 4.2.2.2.as Se reviewer will give our request further consideration.

. 3.4.2.1:

We proposed revision to the Action Statement is related to the Cold Overpressurization Analysis and will receive further consideration as our proposed specifications reflecting the Cold Overpressurization Analysis are reviewed.

4.4.5: .

his specification is undergoing an internal review by the NRC staff. When a resolution has been reached within the staff, our proposed revisions will be addressed.'

4.4.6.2.2:

Se reviewer will have to give our proposal further consideration.

l 3.4.9.2:

There was considerable discussion concerning the 6250F limit which we i proposed for this specification and the 3200F, 10 cycle limit which appears

. in Table 5.7-1. The difference is based on the fact that any time the 6250F limit is exceeded, an evaluation aust be performed whereas the 3200F limit has to be exceeded for more than 10 cycles before an evaluation must be

performed. The concern was addressed by adjusting the limit in Table 5.7-1 to less than 6250F but greater than 3200F for 10 cycles. See Attachment 2

] for the marked-up page.

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. VEGP UNIT 1 IRAFT TECBtICAL SPECIFIC &TIONS MRC SITE VISIT MAY 27-30, 1986 Page 3 3.7.1.1, Action Statements: ,

The reviewer stated that our proposed revisions fell into the category of a short-ters improvement and will be considered with the rest of the short-term improvements.

3/4.7.5:

The reviewer will gi,ve our proposed revisions to this specification further consideration. However, GPC will need to consider operability requirements on the NSCW basin transfer pumps since both basins are required to meet the 30-day cooling requirement. Bechtel will also check on appropriate surveillance requirements for ambient and wet-bulb temperature. Bechtel will also check that 80.25 feet is equivalent to elevation 217'-3".

3/4.7.13:

We asked the reviewer to clarify the basis for this requirement and advise us as to the approach which other utilities have used to delete this requirement. He stated that the basis for the specification on area tesperature monitoring is to ensure that, during normal operation, the equipment qualification limits on temperature are not exceeded. Other utilities have deleted this requirement by demonstrating that the equipment qualification temperature will not be exceeded even given the failure of room

- cooling.

4.5.2.e We agreed to clarify this testing requirement by calling out the RHR suction switchover separately sc that it is clear that actuation is based on RWST low-low level coincident with safety injection.

i 4.6.1.2.c.1:

i GPC will provide further clarification as to the intent of the proposed words "the absolute value of". The concern was that, by adding these words, we are implying that a assative value is possible and this caused some confusion on I

the part of the reviewer.

4.8.2.1.b.2 and 4.8.2.1.c.3: i l

! GPC will provide clarification as to how the average connection resistance and j average cell-to-cell resistance will be determined. In the meantine, the

, specificetions will show "laters" for the limits on connection resistance and cell-to-cell resistance.

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. VBGP !! NIT 1 DRAFT TECENICAL 8FBCIFICATIONS MEC SITE VISIT M&Y 27-30, 1986 Page 9 4.9.1.1 and footnote *:

• s s. .

The reviewer will have to treat our proposed deletion of this surveillance requirement as a short-term improvement.

3.9.8.2, footnote *:

This is also a short-term improvement; however, it has been approved for Callaway and Wolf Creek. The reviewer will give our request further consideration. ,

3.10.3 sad 3.10.4

-GPC will revise the requested revision to clarify that the power range high' serpoint will be iceset at the low setpoint as specified in Table 2.2-1.

, Bases for 4.0.3:

• This will be considered along with our proposed revision to 4.0.3 Bases for 3/4.1.3:

The reviewer will consider the bases proposed by Westinghouse for this specification.

Bases for~3/4.2.3, 3/4.2.4 and 3/4.2.5:

The reviewer will provide the bases for these specifications from the Seabrook Technical Specif.ications.

Bases for 3/4.4.2 and 3/4.4.6.2:

We will provide further clarification for our requested revisions to these bases.

Bases for 3/4.4.9, Page B3/4 4-8:

The reviewer will give our requested revisions to this page further consideration.

Bases for 3/4.4.10: ,

We will provide the correct reference to the ASME Section XI code.

Bases for 3/4.6.1.7:

GPC will consider the Palo Verde bases along with this specification.

. MEETING NOTES

• YEGP UNIT 1 ERAFT TECENICAL SPECIFICATIONS NEC SITE VISIT MAY 27-30. 1986 Page 10 Bases for 3/4.6.2.3:

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Our proposed revisions to these bases resulted in a question concerning the degree of overlap between the cooling capacities for containment spray and the containment coolers. Westinghouse will investigate and provide clarification.

Bases for 3/4.7.8:

We will ensure that the bases are consistent with the specification.

~'5.4.2:

Ihe reviewer will clarify the intent of this specification on RCS volume.

5.6.1.1.at The reviewer will give our prepisal to delete the 2.6% delta K/K further consideration.

SER Review:

The reviewer suggested that we check the following SER sections for impact on the Technical Specifications: 4.4.3.1, 7.1.4.3, 7.3.3.11, 8.4.6, 15.0, and 1.9 of Supplement 1. Attachment 4 is a copy of our SER review which we gave to the reviewer at the meeting.

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Enclosure 3 i

, ADMINISTRATIVE CON 7ROLS i A

6.5.3 TECHNICAL REVIEW AND CONTROL E l ACTIVITIES

• l 6.5.3.1 Activities which affect nuclear safety shall be conducted as follows. '

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a. Procedures required by Specification 6.8 and other procedures which l

affect plant nuclear safety, and changes thereto, shall be prepared, reviewed and approved. Each such procedure or procedure change shall be reviewed by a qualified individual / group other than the

. individual / group which prepared the procedure or procedure change, but who may be from the same organization as the individual / group which prepared the procedure or procedure change. Procedures other l

than Administrative Procedures shall be approved by the appropriate ' ' - ays A vod.

Department Head as .....ggted

., "' ;t,inshall writing by the Administrative approve " - ---

-' 41. ant. The "- 7 7, Procedures, Security Plan implementing procedures and Radiolo'gical Emergency Response Plan implementing procedures. Temporary changes o

to procedures which do not change the intent of the approved proce-dures shall be approved for implementation by two members of the plant staff, at least one of whom holds a Senior Operator license, and documented. The temporary changes shall be approved by the origir.a1 approval authority within 14 days of implementation. For changes to procedures which may involve a change in intent o.f the approved procedures, the person authorized above to approve the N procedure shall approve the change prior to implementation;

, a, e a,o 6

} b. Proposed cha'nges [(modifications to plant nuclear safety-related structures, sys s and components shall be reviewed as designated by the " ..;; r pr=1 -> ; "?::t. Each such modification shall be reviewed by a qualified individual / group other than the individual /

group which designed the modification, but who say be from the same f

organization as the individual / group which designed the modifica-tions. Proposed modifications to plant nuclear safety-related structures, systems and components shall be approved prior to implementation by the ":- r- Ps11-" y Mant;CModj

c. Proposed tests and experiments which affect plant nuclear safety and are not addressed in the Final Safety Analysis Report or Technical Specifications shall be prepareq, reviewed, and approved. Each such test or experiment shall be reviewed by a qualified individual / group other than the individual / group which prepared the proposed test or experiment. Proposed test and experiments shall be approved before implementatton by the "_i;;;r, S? ?; ;, "? " - $ sy4)6) t

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i CALLAWAY - UNIT 1 6-13

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ADMINISTRAllVF CONTROLS

• gijvi,111s(Continued) it. Individu.ils re monsible for reviews performed in icenrdance;with Specifications 6.5.3.la., 6.5.3.lb., and 6.5.3.1c., shall be members of the plant management staff previously designated by the

@ vyd, 5 :;r , C.;;. ., "';nt.. Each such review shall include a deter-mination of whether or not additional, cross-disciplinary, review is necessary. If deemed necessary, such review shall be performed by qualified personnel of the appropriate discipline;

e. Each review shall include a determination of whether or not an unreviewed safety Question is involved. Pursuant to Section 50.59, 10 CFR, NRC approval of items involving unreviewed safety quest 1cns shall be obtained prior to the ".:n;,.., g ... .y "'--*_- approval for implementation; and N

f. The Plant Security Plan and R.u. L,;; Emergency Reepease Plan, and implementing procedures, shall b reviewed at least once per 12 months. Recommended changes to e implementing procedures

- shallbeapprovedbythe"c.....,f:??:r;"?:-'. Recommended l changes to the Plans shall be reviewed pursuant to the requirements of Specifications 6.5.1.6 and 6.5.2.8 and approved by the 44enager,G4s vyd, k'?:n , 7;.nt. NRC approval shall be obtained as appropriate.

RFCOR_DS 6.5.3.2 Records of the abnve activities shall be provided to the Manager, Callaway Plant, ORC and/or NSRB as necessary for required reviews.

j.i. noruxi ABLE EVENT m.iisn -

/e following ctionsshallbetakenforREPORTyLEEVENTS:

l 6.6.1

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a. TheCommissionshal);benotifiedandareport)submittedpursuant to the equirements of Section SOf73 of 10 CFR Part 50, and EactREPORTABLFIVENTshallbereviewedbyth / ORC and submitted to b.

t NSRB and 'e Vice Pres fl t-Nuclear.

6 1 SAFE Y LIMIT VI ATION .

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6.1.1 he followl actions shall he taken in the event a Safety Limit is viol ed:

a. lhe NC Operations C ter shall be notifi d by telephone as soon as po ible and in all cases within I hour / The Vice President-Nuclear a d the NSRB shal be notified within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />;

b. A Safety limit iolation Repor', sha 1 be prepared. The report shall be reviewed py the ORC. This reps'rt shall describe: (1) applicable citetsastances preceding the vio dtion; (2) effects of the violation upon facilit components, syst 5 or structures; and (3) corrective action taken o prevent recur ence; }

l . . . . tu,l' 1 6-14 i

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Enclosure 4 SECTION 6.0 ADMINISTRATIVE CONTROLS O

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i ADMINISTRATIVE CONTROL 5 _. ,

6.1 RESPONSIBILITY 1 l

6.1.1 The General Manager - Vogtle Nuclear Operati8ns (GMVNO) shall be respon-e sible for overall plant operation and shall delegate in writing the succession to this responsibility during his absence. -

6.1.2 he S ft Su)ervisor (or during his absence from the control room, 1

n ag C f c o ofate lanageM nt] shall be reissued to all station __ ,personnel , 4 ,, on g an I

nualL4 7A g_g 7 g 6.2 ORGANIZATION =y 1AL ' ' 4 A". a Np* 46 Ano.~~, &

'*^=eM W = " : /4 ===Af = m_j k M "" dsi Q OFFSITE The offsite organization for plant

.a. s .-

management iD Y _

and technical support 6.2.1 shall be as shown in Figure 6.2-1.

UNIT STAFF

6.2.2 The unit organization shall be as shown in Figure 6.2-2 and:

a. Each on-duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1;

b. At least one licensed Operator shall be in the control room when fuel is in the reactor. In addition, while the unit is in MODE 1, 2, 3, or 4, at least one licensed Senior Operator shall be in the control room;

c. A Health Physics Technician

• shall be on site when fuel is in the reactor;

d. All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation;

e. A site Fire Brigade of at least five members

• shall be maintained on site at all times. The Fire Brigade shall not include the Shift Supervisor and the [two) other members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency; and

• The Health Physics Technician and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to accommodate unexpected absence, provided immediate action is taken to fill the required positions.

6-1 V0GTLE - UNIT 1 APR 24 MISC l

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, ADMINISTRATIVE CONTROLS UNIT STAFF (Continued)

A f.

Administrative procedures shall be developed and implemented to limit 1 the working hours of plant staff in performance of safety-related functions (e.g. , licensed Senior Operators, licensed Operators, key HealthPhysicsTechnicians,/eynon-licensedoperators,andkey maintenance personnel).

l or Adequate shift coverage shall be maintained without routine heavy use of overtin The objective shall be to have operating personnel l' work a a:Tfg:r ty, 40-hour week while the plant is operating.c' However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shut-down for refueling, major maintenance, or major plant modification, ed on a tempor ry basis the following guidelines shall be. folloy M: 3'

/D M M WVa M MM An individual should not be permitted to work more than 16 hcurs 1.

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-hour 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-hour 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.

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 work periods, including shift turnover time.

4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the

- entire staff on a shift.

Any deviation frem the above guidelines shall be authorized by the g4/A yf [Phat S perktdy;at] ;r 7,io /,.pdf, or higher levels of manage-9_gxj W ment, in accordars:e with established procedures and with documenta-tion of'the basiR for granting the deviation. Controls shall be included in the paocedures such that individual overtime shall be reviewed monthly by the General Manager - Vogtie Nuclear Operations or his designee to assure that excessive hours have not been assigned.

Routine deviation from the above guidelines i:; not authorized.

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V0GTLE - UNIT 1 6-2 APR 24 N 1 ,

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FIGURE 6.2-1 0FFSITE ORGANIZATION V0GTLE - UNIT 1 6-3 t

RAFT O

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FIGURE 6.2-2 UNIT ORGANIZATION V0GTLE - UNIT 1 6-4 APR 241986 l

l TABLE 6.2-1 .

MINIMUM SHIFT CREW COMPOSITION SINGLE UNIT FACILITY e

POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODE 1, 2, 3, or 4 MODE 5 or 6 1 1 05 SRO 1 None R0 2 1 NLO 2 1 1* None l STA ,

OS - Operations Supervisor with a Senior Operator license on Unit 1 SRO - Individual with a Senior Operator license on Unit 1 RO - Individual with an Operator license on Unit 1 NLO - Non-Licensed Operator STA Shift Technical Advisor The shift crew composition may be one less than the minimum requirements of Table 6.2-1 for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements

, of Table 6.2-1. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

During any absence of the Shift Supervisor from the control room while the unit is in MODE 1, 2, 3, or 4, an individual (other than the Shift Technical Advisor) with a valid Senior Operator license shall be designated to assume the control room command function. During any absence of the Shift Supervisor from the control room while the unit is in MODE 5 or 6, an individual with a valid Senior Operator license or Operator license shall be designated to assume the control room command function.

• The STA position shall be manned in MODES 1, 2, 3, and 4 unless the Shift Supervisor or the individual with a Senior Operator license meets the l

qualifications for the STA as required by the NRC.

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l V0GTLE - UNIT 1 6-5 APR 84 M

ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

FUNCTION 6.2.3.1 The ISEG shall function to examine plant operating characteristics, NRC issuances, industry advisories, Licensee Event Reports, and other sources of plant design and operating experience information, 4W"Mac =it: Of :f-Mr dadon; which may indicate areas for improving plant safety. The ISEG shall make detailed recommendations for revised procedures, equipment modifications, maintenance activities, operations activities, or other means of improving plant safety to the Senior Vice President-Nuclear Operations through the Manager-Nuclear Performance and Analysis.

COMPOSITION .'.

6.2.3.2 The ISEG shall be composed of at least five, dedicated, full-time engineers. W t d = :ite. Each shall have a bachelor's degree in engineering or related science and at least 2 years professional level experience in his field, at least 1 year of which experience shall be in the nuclear field.

RESPONSIBILITIES 6.2.3.3 The ISEG shall be responsible for maintaining surveillan:e of unit activities to provide independent verification

• that these activities are performed correctly and that human errors are reduced as much as practical.

RECORDS

- 6.2.3.4 Records of activities performed by the ISEG shall be prepared, main-tained, and forwarded each calendar month to [a high level corporate official in a technically oriented position who is not in the management chain for powerproduction].

6.2.4 SHIFT TECHNICAL ADVISOR

! 6.2.4.1 The Shift Technical Advisor shall provide advisory technical support to the Shift Supervisor in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. The Shift Technical Advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline and shall have received specific training in the response and analysis of the unit for transients and accidents, and in unit design and layout, including the capabilities of instrumentation and controls in the control room.

6.3 UNIT STAFF QUALIFICATIONS f /'W Q4 pyg. C. T-l ppm NM

• Not responsible for sign-off function.

V0GTLE - UNIT 1 6-6 gg

ADMINISTRATIVE CONTROLS UNIT STAFF QUALIFICATIONS (Continued) 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifica-tions of ANSI /ANS N18.1971 for comparable positions, except for the [ Radiation Protection Manager] who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975. The licensed Operators and Senior Operators shall also meet or exceed the minin.um qualifications of the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees.

6.4 TRAINING 6.4.1 A retraining and replacement training program for the plant staff shall be maintained under,the direction of the Superintendent of Nuclear Training and shall meet or exceed the requirements and reconenendations of Section of ANSI /ANS N18.1971 and Appendix A of 10 CFR Part 55 and the supplement W require-ments specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational expe.rience.

6.5 REVIEW AND AUDIT 6.5.1 PLANT REVIEW BOARD (PRB)

FUNCTION 6.5.1.1 The PRB shall function to advise the GMVNO on all matters related to nuclear safety.

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COMPOSITION 6.5.1.2 The PRB shall be composed of the:

Chairman: [ Plant Superintendent]

Member: [ Operations Supervisor)

Member: [ Technical Supervisor]

Member: [ Maintenance Supervisor]

Member: [ Plant Instrument and Control Engineer]

Member: [ Plant Nuclear Engineer]

Member: [ Health Physicist]

ALTERNATES l 6.5.1.3 All alternate members shall be appointed in writing by the PR8 Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in PRB activities at any one time.

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V0GTLE - UNIT 1 6-7 l APR 2412 j

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- ADMINISTRATIVE CONTROLS I i

MEETING FREQUENCY l

~6.5.1.4 The PRS shall meet at least once per calendar month and as convened by the PRS Chairman or his designated alternate.

QUORUM  !

6.5.1.5 The quorum of the PRB necessary for the performance of the PR8 responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and four members including alternates..

gP gol RESPONSIBILITIES 6 ^#

g i 4 6.5.1.6 The PRS shall be responsible for:

a. Review of: (1) all proposed procedures required by Specification 6.8 and changes thereto, (2) all proposed programs required by Specification 6.8 and changes thereto, and (3) any other proposed procedures or changes thereto as determined by the GMVNO to affect nuclear safety;

! . b. Review of all proposed tests and experiments that affect nuclear safety; f

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' Review of all proposed changes to the Technical Specifications; c.

d. Review of all proposed changes or modifications to unit systems or equipment that affect nuclear safety;

e. Investigation of all violations of the Technical Specifications, including the preparation and forwarding of reports covering evalua-tion and recommendations to prevent recurrence, to the Vice President and General Manager-Nuclear Operations and to the Safety Review Board;

f. Review of all REPORTABLE EVENTS; l

1. g. Review of plant operations to detect potential hazards to nuclear

• safety; 1

h. Performance of special reviews, investigations, or analyses and reports thereon as requested by the GMVNO or the Safety Review Board; i.

Review of the Security Plan and implementing procedures and submittal of recommended changes to the GMVNO and the Safety Review Board; i

i V0GTLE - UNIT 1 6-8 APRS4 m  !

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ADMINISTRATIVE CONTROLS l

RESPONSIBILITIES (Continued)

j. Review of the Emergency Plan and implementing procedures and submittal of recommended changes to the GMVNO and the Safety Review Board;

k. Review of any accidental, unplanned, or uncontrolled radioactive release including the preparation of reports covering evaluation, recommendations, and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vice President and General Manager-Nuclear Operations and to the Safety Review Board; and 1.I' Review of changes to the PROCESS CONTROL PROGRAM, the OFFSITE DOSE l I

CALCULATION MANUAL, and the Radwaste Treatment Systems.

m. Review of the Fire Protection Program and Implementing procedures and submittal of recommended changes to the GMVNO.

6.5.1.7 The PRB shall:

a. Recommend in writing to the GMVNO approval or disapproval of items considered under Specification 6.5.1.6a. through d. prior to their implementation;

b. Render determinations in writing with regard to whether or not each item considered under Specification 6.5.1.6a. through e. constitutes an unreviewed safety question; and

c. Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Vice President and General Manager-Nuclear Operations and the Safety Review Board of disagreen. ant between the PRB and the GMVN0; however, the GMVNO

- shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1.

RECORDS 6.5.1.8 The PRB shall maintain written minutes of each PRB meeting that, at a minimum, document the results of all PRB activities performed under the respon-l

' sibility provisions of these Technical Specifications. Copies shall be provided

[

to the Vice President and General Manager-Nuclear Operations and the Safety Review Board.

6.5.2 SAFETY REVIEW BOARD (SRB)

FUNCTION 6.5.2.1 The SRB shall function to provide independent review and audit of designated activities in the areas of:

a. Nuclear power plant operations, I b. Nuclear engineering,

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V0GTLE - UNIT 1 6-9 APR84IIEi

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ADMINISTRATIVE CONTROLS

- FUNCTION (Continued)

c. Chemistry and radiochemistry,

d. Metallurgy,

e. Instrumentation and control,

f. Radiological safety,

g. Mechanical and electrical engineering, h.., Quality assurance practices, and The SRB shall report to and advise the Senior Vice President-Nuclear Operations on those areas of responsibility specified in Specifications 6.5.2.7 and 6.5.2.8.

COMPOSITION 6.5.2.2 The SRB shall be composed of the:

Director: [ Position Title]

Member: [ Position Title)

Member: [ Position Title]

Member: [ Position Title]

Member: [ Position Title]

ALTERNATES

- 6.5.2.3 All alternate members shall be appointed in writing by the SRB Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in SRB activities at any one time.

CONSULTANTS

. 6.5.2.4 Consultants shall be utilized as determined by the SRB chairman to provide expert advice to the SRB.

MEETING FREQUENCY 6.5.2.5 The SRB shall meet at least once per calendar quarter during the initial year of plant operation following fuel loading and at least once per 6 months thereafter.

l V0GTLE - UNIT 1 6-10 A9R 34 %

ADMINISTRATIVE C0aTROL$

!* 000 RUM 6.5.2.6 The quorum of the SRB necessary for the performance of the SRB review and audit functions of these Technical Specifications shall consist of the Chairman or his designated alternate and at least four SRB members including alternates. No more than a minority of the quorum shall have line responsibility for operation of the plant.

REVIEW 6.5.2.7 The SRB shall be responsible for the review of:

af ' The safety evaluations for: (1) changes to procedures, equipment, or systems; and (2) tests or experiments completed under the provision of 10 CFR 50.59, to verify that such actions did not constitute an unreviewed safety question;

b. Proposed changes to procedures, equipment, or systems which involve an unreviewed safety question as defined in 10 CFR 50.59;

c. Proposed tests or experiments which involve an unreviewed safety question as defined in 10 CFR 50.59;

d. Proposed changes to Technical Specifications or this Operating License;

e. Violations of Codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance;

f. Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety;

g. All REPORTA8LE EVENTS;

h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety; and s

Reports and meeting minutes of the PRB.

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j AUDITS 6.5.2.8 Audits of plant activities shall be performed under the cognizance of the SRB. These audits shall encompass:

a. The conformance of plant operation to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months; V0GTLE - UNIT 1 6-11

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I ADMINISTRATIVE CONTROLS

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AUDITS (Continued)

. b. The performance, training, and qualifications of the entire plant staff at least once per 12 months;

c. The results of actions taken to correct deficiencies occurring in -

plant equipment, structures, systems, or method operation that affect nuclear safety, at least once per 6 months;

d. The performance of activities required by the Operational Quality Assurance Program to meet the criteri& of Appendix 8, 10 CFR Part 50, at least once per 24 months;

, e.: The fire protection programmatic controls including the implementing procedures at least once per 24 months by qualified licensee QA personnel;

. f. The fire protection equipment and program implementation at least once per 12 months utilizing either a qualified offsite licensee fire protection engineer or an outside independent fire protection consultant. An outside independent fire protection consultant shall be used as least every third year;

g. The Radiological Environmental Monitoring Program and the results thereof at least once per 12 months;

h. The 0FFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months; The PROCESS CONTROL PROGRAM and implementing procedures for processing l '-

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and packaging of radioactive wastes at least once per 24 months;

j. The performance of activities required by the Quality Assurance

! Program for effluent and environmental monitoring at least once per l 12 months;

- k. The Emergency Plan and implementing procedures (at least once per 12 months);

1. The Security Plan and implementing procedures (at least once per 12 months); and q ther area of unit operation considered appropriate by the SRB or the Vice President-Nuclear Operations.

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RECORDS 6.5.2.9 Records of SRB activities shall be prepared, approved, and distributed as indicated below:

a. Minutes of each SRB meeting shall be p'repared, approved, and fo marded to the Senior Vice President-Nuclear Operations within 14 days follow-ing each meeting; 1

b. Reports of reviews encompassed by Specification 6.5.2.7 shall be pre-pared, approved, and fomarded to the Senior Vice President-Nuclear Operations within 14 days following completion of the review; and V0GTLE - UNIT 1 6-12 APR34 W6 e

N ADMINISTRATIVE CONTROLS 1 RECORDS (Continu;d)

c. Audit reports encompassed by Specification 6.5.2.8 shall be forwarded to the Executive Vice President-Power Supply, Senior Vice President-Nuclear Operations and to the management positions responsible for the areas audited within 30 days after completion of the audit by the

! auditing organization.

6. 6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTA8LE EVENTS:

a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.72 and Section 50.73 to 10 CFR Part 50, and

b. Each REPORTA8LE EVENT shall be reviewed by the PRB, and the results of this review shall be submitted to the SRB and the Vice President j

and General Man'ager-Nuclear Operations.

6.7 SAFETY LIMIT VIOLATION 6.7.1 The following action,s shall be taken in the event a Safety Limit is violated:

, a. In accordance with 10 CFR 50.72, the NRC Operations Center shall be

' notified by telephone as soon as practical and in all cases within one hour after the violation has been determined. The Vice

! President and General Manager-Nuclear Operations, the SRB, PRB, and the GMVNO shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. A Licensee Event Report shall be prepared in accordance with

. 10 CFR 50.73.

c. The Licensee Event Report shall be submitted to the Commission in accordan:e with 10 CFR 50.73, and to the PRB, SRB, the GMVNO and the Vice President and General h3 nager-Nuclear Operations within 30 days after discovery of the event,

d. Critical operation of the unit shall not be resumed until authorized by the Nuclear Regulatory Commission.

' 6.8 PROCEDURES AND PROGRAMS i

6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:

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

b. The emergency operating procedures required to implement the require-ments of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Generic Letter No. 82-33; V0GTLE - UNIT 1 6-13 APR 84 Wes

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IMAtl ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

c. Security Plan implementation;

d. Emergency Plan implementation;

e. PROCESS CONTROL PROGRAM implementation; f.

OFFSITEDOSECALCUlhTIONMANUALimplementation;and

g. Quality Assurance for effluent and environmental monitoring.

h. ~' Fire Protection Program Implementation.

p 6.8.2 Each procedure of Specification 6.8.1, and changes thereto, shall be reviewed by the PRB and shall be approved by the GMVNO prior to implementation [p and reviewed periodically as set forth in administrative procedures.

6.8.3 Temporary changes to procedures of Specification 6.8.1 may be made pro-vided:

a. The intent of the original procedure is not altered; N

b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Operator license; and Q$ %

The change is documented, reviewed by the PRB, and approved by the y

c.

GMVNO within 14 days of implementation.-

6.8.4 The following programs shall be established, implemented, and maintained:

a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids duringThea serious' transient or accident to as low as practical levels.

systems include the applicable portions of containment recirculation spray, Safety Injection, chemical and volume control, and residual heat removal systems. The program shall include the following:

1) Preventive maintenance and periodic visual inspection requirements, and

2) 4 e r =+=A.M ak test requirements for each system at refueling cycle intervals or less.

b. In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

! V0GTLE - UNIT 1 6-14 APR 2 e 1986 l

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continuid) l

b. In-Plant Radiation Monitorina (Continued)

1) Training of personnel,

2) Procedures for monitoring, and

3) Provisions for maintenance of sampling and analysis equipment.

c. Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shall include:

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1) Identification of'a sampling schedule for the critical variables and control points for these variables,

2) Identification of the procedures used to measure the. values of the critical variables, .,

3) Identification of process sampling points, which shall include j. . .-

monitoring the discharge of the condensate pumps for evidence of condenser in-leakage,

4) Procedures for the recording and management of data,

5) Procedures defining corrective actions for all off-contr,o1 point chemistry conditions, and

_ 6) A procedure identifying: (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective action.

d. Post-Accident Sampling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant

! gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include the following:

1) Training of personnel, j

2) Procedures for sampling and analysis, and

3) Provisions for maintenance of sampling and analysis equipment.

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l l ADMINISTRATIVE CONTROLS l

6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless othemise noted.

STARTUP REPORT 6.9.1.1 A summary report (1) of plant startup and power escalation testing shall receipt of an Operating License, (2) amendment to be submitted following:

the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit.

The initial Startup Report shall address each of the startup tests iden-tified in the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satis-factory 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/or modifications.

Startup Reports shall be submitted within: (1) 90 days following completion of the Startup Test Program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If 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 l

shall be submitted at least every 3 months until all three events have been l

completed.

l ANNUAL REPORTS 6.9.1.2 Annual Reports covering the activities of the plant as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality.

Reports required on an annual basis shall include:

a. A tabulation on an annual basis of the number of plant, utility, and other personnel (including contractors) receiving exposures greater than 100 mres/yr and their associated man-res exposure according to work and job functions *.(e.g., reactor operations and "This tabulation supplements the requirements of 520.407 of 10 CFR Part 20.

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ADMINISTRATIVE CONTROLS 4

ANNUAL REPORTS (Continued)

- surveillance, inservice inspection, routine maintenance, special main-tenance [ describe maintenance), waste processing, and refueling). The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD), or film badge )

measurements. Small exposures totalling less than 20% of the individual

- total dose need not be accounted for. In the aggregate, at least 80% l of the total whole-body dose received from external sources should be  !

assigned to specific major work functions;

b. The results of specific activity analyses in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be' included: (1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded (in graphic and tabular format); (2) Results of the last isotopic analysis for radio-iodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiotodine activity was reduced to less than limit. Each result should include date and time of sampling and the radiciodine concentrations; ,

(3) Clean-up flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration (pCi/gn) and one other radioidine isotope concentration (pCi/gs) as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.

c. A report shall be prepared and submitted to the commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcuries of removable contamination.

d. An annual data report on diesel generator reliability will be submitted and, in addition, the following information will be included:

l< A summary of all tests (valid and invalid) that occurred within 1 1.

the time [ period over which the last 20/100 valid tests were

> performed].

2. Analysis of failures and determination of root causes of failures.

f 3. Identification of all actions taken or to be taken to 1) correct the root causes of failures defined in b) above and 2) achieve a general improvement of diesel generator reliability.

l

4. An assessment of the existing reliability of electric power to engineered-safety-feature equipment.

ANNUAL RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE REPORT _-

6.9.1.3 Routine Annual Radiological Environmental Surveillance Reports covermgr4- &

tM r;r : tie- ? t" "-M during the previous calendar year shall be submitted

(

prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following initial criticality and shall include copies of reports of the preoperational Radiological Environmental Monitoring Program of l' the unit for,at least two years prior to initial criticality.

l i

V0GTLE - UNIT 1 6-17 h9R S 4 M l

l l

- - , . , , - - - - - .w _. _ - - , _ . - - - - , . . . - , - - , , , , - - - - - - . _ - _ _.. _.-.- --- ___ . .- ..__._

ammmmm_ . ___

y,gya m ^~^S'"sh W ANNUAL RADIOLOGICAL ENVIRONMENTAL 9PEAA m e REPORT (Continued)

The Annual Radiological Environmental Surveillance Reports shall include summaries, interpretations, and an analysis of trends of the results of the j radiological environmental surveillance activities for the report period, including,a comparison with preoperatic..a1 studies, with operational controls, (Ts appropriaty, and with previous environmental surveillance reports, and an assessment. of the observed impacts of the plant operation on the environment. {

The reports shall also include the results of the Land Use Census required by Specification 3.12.2.

The Annual Radiological Environmental Surveillance Reports shall include the resuits of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the Offsite Dose Calculation Manual, as well as summarized and tabulated results of these analyses and measurements in the format'of the table in the Radiological Assessment Branch Technical Position, Revision 1. November 1979. G n the event that some inoivi-dual results are not available for inclusion with the report, the report shall The be submitted noting and explaining the reasons for the missing results.

miss g d g a 1 p submitted as soon as possible in a supplementary rep

• ihereportsshallalsoincludethefollowing: a summary description of 1-- d M legible maps

• the Radiological Environmental Monitoring Program; **

vering all sampling locations keyed to a table giving distances and directions from 3- cen+aW .; ;' r: reactor); the results of licensee participation in the Interlaboratory Comparison Program and the corrective action taken if the

- specified program is not being performed as required by Specification 3.12.3; reasons for not conducting the Radiological Environmental Monitoring Program as required by specification 3.12.1, and discussion of all deviations from the ments that exceed the reporting levels of Table 3.12-2 but are n of plant effluents, pursuant to ACTION b. of Specification 3.12.1; and discussion of all analyses in which the LLO required by Table 4.12-1 was not achievable.

SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT l 6.9.1.4 Routine Semiannual Radioactive Effluent Release Reports covering the

operation of the. unit during the previous 6 months of operation shall beThe period submitted within 60 days after January 1 and July 1 of each year.

of the first report shall begin with the date of initial criticality.

The Semiannual Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21

~

" Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data

• 0ne map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations.

V0GTLE - UNIT 1 6-18 '

APR S4 W 6 f

SENIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued) summarized on c quarterly basis following the format of App 2ndix 8 thereof.

For solid wastes, the format for Table 3 in Appendix B shall be supplemented with three additional categories: class of solid wastes (as defined by 10 CFR Part 61), type of container (e.g., LSA, Type A, Type B, Large Quantity) and SOLIDIFICATION agent or absorbent (e.g., cement, urea formaldehyde).

The Semiannual Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary This annual summary of hourly may meteorological data collected over the previous year.

be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured),

or in the form of joint frequency distributions of wind speed, wind direction, )

and atmospheric stability.* This same report shall include an assessment of I the radiation doses due to the radioactive liquid and gaseousThis effluents same report released from the unit or station during the previous calendar year. '

shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figure [5.1-1]) during the report period. All l

assumptions used in making these assessments, i.e., specific activity, exposure time, and location, shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall .

be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (00CM).

The Semiannual Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating j .

the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev.1, October 1977.

The Semiannual Radioactive Effluent Release Reports shall include a list

> and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (00CM), pursuant to Specifica-tions 6.13 and 6.14, respectively, as well as any major change to Liquid,

• In lieu of submission with the Semiannual Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required ,

meteorological data on site in a file that shall be provided to the NRC upon request.

6-19 gpg g 4 gg V0GTLE - UNIT 2 f

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

ADMINISTRATIVE CONTROLS SENIANNUAL RADIDACTIVE EFFLUENT RELEASE REPORT (Continurd)

Gaseous, or Solid Radwaste Treatment Systems pursuant environmental monitoring identified by the Land Use Census pursuant to Speci-fication 3.12.2.

The Semiannual Radioactive Effluent Release Reports shall also include the following: an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Specification 3.3.3.10 or 3.3.3.11, respectively; and description of the events Teading to liquid holdup tanks or gas storage tanks exceeding the limits of Specificat' ion 3.11.1.4 or 3.11.2.6, respectively.

MONTHLY OPERATING REPORTS 6.9.1.5 Routine reports of' operating statistics and shutdown experience, '

including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the Director, Office of Resource -

Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a i

copy to the Regional Administrator of the Regional Office of RADIAL PEAKING FACTOR LIMIT REPORT

6. 9.1. 6 The F,y limits for RATED THERMAL POWER (F P) shall be pro the NRC Regional Administrator with a copy to Director of Nuclear Re Regulation, Attention: 20555, for all core planes containing Bank "0" Commission, Washington, D. C.

r control rods and all unrodded core planes and the plot of predicted (F Pg ,j) l vs Axial Core Neight with the limit envelope at least 60 days prior to each l'

cycle initial criticality unless otherwise approved by the Commission by letter.

In addition, in the event that the limit should change requiring a new substan-tial or an amended submittal to the Radial Peaking Factor Limit Report, it will be submitted 60 days prior to the date the limit would become effective unless otherwise approved by the Commission by letter. Any information needed to support F will be by request from the NRC and need not be included in this report.

4 6-20 g V0GTLE - UNIT 1 1

N ADMINISTRATIVE CONTROLS SPECIAL REPORTS i

6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

6.10 RECORD RETENTION f

6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.2 The following records shall be retained for at least 5 years:

a. Records and logs of plant operation covering time interval 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. All REPORTABLE EVENTS;

d. Records of surveillance activities, inspections, and calibrations required by these Technical Specifications;

e. Records of changes made to the procedures required by Specification 6.8.1;

f. Records of radioactive shipments;

g. Records of sealed source and fission detector leak tests and results; and

h. Records of annual physical inventory of all sealed source material j of record.

? 9,'

i

~

6-21 APR 84 586 V0GTLE - UNIT 1 l

I r

L'~ClLLRJ ADMINISTRATIVE CONTROLS RECORD RETENTION (Continued) 6.10.3 The following records shall be retained for the duration of the plant Operating License:

a. Records and drawing changes reflecting plant design modifications made to systems and equipment described in the Final Safety Analysis Report;

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;

d. Records of gaseous and liquid radioactive material released to the environs;

e. Records of transient or operational cycles for those plant components 1

identified in Table 5.7-1;

f. Records of reactor tests and experiments;

g. Records of training and qualification for current members of the plant staff;

h. Records of inservice inspections performed pursuant to these Technical Specifications;

i. Records of quality assurance activities required by the Operational

- Quality Assurance Manual;

j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59;

k. Records of meetings of the PRB and the SR8; I

1. Records of the service lives of all hydraulic and mechanical i

snubbers required by Specification 3.7.9 including the date at which the service life commences and associated installation and

- maintenance records;

" m. Records of secondary water sampling and water quality; and

n. Records of analyses required by the Radiological Environmental Monitoring Program that would permit evaluation of the accuracy of the analysis at a later date. This should include procedures ,

effective at specified times and QA records showing that these

! procedures were followed.

6.11 RADIATION PROTECTION PROGRAM 6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and i'

adhered to for all operations involving personnel radiation exposure.

l YOGTLE - UNIT 1 6-22 ggg I

i

ADMINISTRATIVE CONTROLS 6.12 HIGH RADIATION AREA [0PTIONAL]

6.12.1 Pursuant to paragraph 20.203(c)(5) of 10 CFR Part 20, in Ifeu of the l

" control device" or " alarm signal" required by paragraph 20.203(c),each high radiation area, as defined in 10 CFR Part 20, in which the intensity of radia-tion is equal to or less than 1000 mR/h at 45 cm (18 in.) from the radiation source or from any surface which the radiation penetrates shall be barricaded l c j 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 Technician) l or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radia-tion areas with exposure rates equal to or less than 1000 mR/h, 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 areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device which continuously indicates the radiation dose rate in the area; or

b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms 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 have been made knowledgeable of them; or

c. An individual qualified in radiation protection procedures with a

' - radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the (Radiation Protection Manager) in the RWP.

6.12.2 In addition to the requirements of Specification 6.12.1, areas accessible

- to personnel with radiation levels greater than 1000 mR/h at 45 cm (18 in.)

from the radiation source or from any surface which the radiation penetrates l

shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative Doors control shall remain of thelocked shift Foreman except during on duty and/or health physics supervision.

periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work areas and the maximum allowable stay In lieu of the stay time specification of .

, time for individuals in that area. ,

the RWP, direct or remote (such as closed circuit TV cameras) continuous  ! A( '

' surveillance may be made by personnel qualified in radiation protection N procedures to provide positive exposure control over the activities being [- l performed within the area. -

1 l

For individual high radiation areas accessible to personnel with radiation -

1evels of greater than 1000 mR/h that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and where l no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded, conspicuously posted, and a flashing light shall be activiated as a warning device.

V0GTLE - UNIT 1 6-23 APR 34 96 1

1

. ADMINISTRATIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 The PCP shall be approved by the Commission prior to implementation.

6.13.2 Licensee-initiated changes to the PCP:

a. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made. This subnittal shall contain:

1) Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;

2) A deter:nination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and

3) Documentation of the fact that the change has been reviewed and found acceptable by the PRB.

b. Shall become effective upon review and acceptance by the PRB.

6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 The ODCM shall be approved by the Commission prior to implementation.

6.14.2 Licensee initiated changes to the ODCH:

a. Shall be submitted to the Comission in the Semiannual Radioactive Effluent Release Report for the period ir. which the change (s) was made effective. This submittal shall contain:

1) Sufficiently detailed information to totally support the rationale for the change without benefit of additional or

! supplemental information. Information submitted should consist of a package of those pages of the 00CM to be changed with each page numbered, dated and containing the revision number, together

. with appropriate analyses or evaluations justifying the change (s); .

2) A determination that the change will not reduce the accuracy or reliability of dose calculations or Setpoint determinations; and

3) Documentation of the fact that the change has been reviewed and found acceptable by the PRB.

b. Shall become effective upon review and acceptance by the PRB.

i V0GTLE - UNIT 1 6-24 hPR S 4 %

, 'ADMkNI5TRATIVECONTROLS i

4 i

6.15 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLIO RADWASTE TREATMENT SYSTEMS * ,

1 6.15.1 Licensee-initiated major changes to the Radwaste Treatment Systems (liquid, gaseous, and solid):

a. Shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the PRB. The discussion of each change shall contain:

1) A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59;

2) Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;

3) A detailed description of the equipment, components, and processes involved a'nd the interfaces with other plant systems;

4) An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous affluents and/or quantity of solid waste that differ from those previously predicted in the License application and amendments thereto;

5) An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the License application and amendments thereto;

6) A comparison of the predicted releases of radioactive materials,

- in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the change is to be made;

7) An estimate of the exposure to plant operating personnel as a result of the change; and

8) Documentation of the fact that the change was reviewed and fcund acceptable by the PRB.

b. Shall become effective upon review and acceptance by the PRB.

• Licensees may choose to submit the information called for in this Specification as part of the annual FSAR update.

V0GTLE - UNIT 1 6-25 APR 84 W6 l

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Enclosure 5

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/ssp E88gg UNITED STATES NUCLEAR REGULATORY COMMISSION g Enclosure 6

! o f y, j wasHWGToN, D. C. 20555

.,- s s *"** j NAY 2 0 586 i Docket No. 50-443 Mr. Robert J. Harrison 6 President & Chief Executive Officer Public Service Company of New Hampshire Post Office Box 330 Manchester, New Hampshire 03105

SUBJECT:

SEABROOK STATION TECHNICAL SPECIFICATION IMPROVEMENT PROGRAM In your May 29, 1985 letter to the staff, you described a Seabrook Station' Technical Specification Improvement Program which you were developing in support of proposed improvements over the Standard Technical Specifications (STS) for the Technical Specifications to be issued with the operating license for Seabrook. The specific changes implementing your program were later merged with other plant-specific changes to the STS. You submitted these proposed Technical Specifications on July 26, 1985 for staff review. The staff has been reviewing your proposed Technical Specifications since that i -

. time and has given careful consideration to your proposed improvements to the STS. Many of the goals of your program coincide with the eventual goals 4

,of our ongoing Technical Specification Improvement Program.

Our review to date of your proposals and the results are described in Enclosure 1.

In summary, we have accepted some of your recommended changes including the transfer of several issues, traditionally included in the plant's Technical Specifications, to other documents. The staff's review of this transfer pro-1 posal is discussed in Enclosure 2. Other recommendations for improvement have i been modified from your proposal. Other proposed changes have not been included in the Seabrook Technical Specifications. This is based upon the staff's review of the information and justifications you provided. Some of the improvements

= not accepted may be covered in the staff's ongoing improvement program and, as a result, are not ready for implementation on operating reactors at this time.

If you believe additional rationale exists for further review, we welcome your submittal of additional technical bases in support of your proposed changes.

We plan to incorporate the changes in Enclosure 1 in the Seabrook Station l Technical Specifications, subject to the transfer of these requirements to I other documents (as discussed in Enclosure 2), as indicated and necessary.

We will work with you in developing the required documentation, including provisions for control of changes to these documents in the future. As necessary, the operating license may include conditions related to control of changes to documents.

l l

1

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'l Mr. Robert J. Harrison EAY 2 01986 1 I realize that your program represents a sizable investment of resources and is a commendable effort. I support your efforts in improving the Seabrook Station Technical Specifications and commit to review additional bases you may submit for existing proposals. We will also consider other si'gnificant improvements you may propose in the future.

s Thomas M. Novak, Acting Director Division of PWR Licensing-A Office of Nuclear Reactor Regulation

Enclosures:

1. Seabrook Technical Specification Improvements Summary

2. Seabrook Technical Specification Improvement Procedures S

l I

- , - - - , n---.. - , , , . . , _ . , , . . , _ , - - - , _ , , . . , . , _ , , . _ , , . _ . _ _ . . , . _ , _ - . - - - - - . - . . - - . . . _ . . . . -, - - , - - - - - _ _ , - , . -

Enclosure 1

- Seabrook Technical Specification Improvements Summary A. Deletion of Technical Specifications .

l 3/4.3.3.8 LOOSE PART DETECTION INSTRUMENTATION ,

The staff position is that Technical Specification 3/4.3.3.8 can be deleted. Because the required ACTION is only a reporting requirement, the whole Technical Specification can be eliminated provided the reporting and surveillance requirements are listed in the FSAR.

3/4.3.4 TURBINE OVERSPEED PROTECTION SYSTEM l

The staff position is that Technical Specification 3/4.3.4

' Turbine Overspeed Protection System be retained in the Seabrook Technical Specifications. The staff's review of the probabilistic

- arguments presented by the applicant did not confirm that the Safety

' Evaluation acceptance criterion (probability of turbine missiles

-5 less than 10 per year) could be met without the Limiting Condition l of Operation on the operability of th"e turbine stop and governor l l

valves. Further, the request for relaxed surveillance schedules I on the turbine stop and governor valves are not supported by the staff's review of the probabilistic assessment nor the manufacturer's reccesMndations.

2 3/4.4.5 STEAM GENERATORS The staff position is that Technical Specification 3/4.4.5 STEAM GENERATORS be retained in the Seabrook Technical Specifications.

The reason is that the applicant proposes to rely on the ISI/IST program which we understand is not scheduled for submittal for' staff review until after issuance of an operating license. However,

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

-1

, the primary reasons for rejecting the Deletion of the Technical Specification are that (1) the specification is important to the maintenance of the primary coolant pressure boundary integrity and (2) the surveillance includes reporting requirenegts that are l not derived from Section XI of the ASME Code and that are not included in the ISI program.

6.3 UNIT STAFF QUALIFICATIONS  ;

l The applicent proposes to delete Specification 6.3, Station Staff Qualification. The staff finds this proposal acceptable because the staff's Safety Evaluation includes a finding of ecceptable

! criteria to be used by the applicant, and because changes to these criteria under the provisions of 10 CFR 50.59 will afford an adequate opportunity for review by the staff.

B. Deletion of Limitina Conditions for Ooeration 3.4.6.2(e). 52 gpa CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 220 psig.

l In the July 26, 1985 submittal, the applicant proposed to delete this LC0 and the corresponding surveillance requirement. The STS

! Bases state. that the controlled leakage limitation ensures that l in the event of a LOCA, the safety injection flow will not be less than assumed in the accident analysis. In his comments on the proof and review version of the Technical Specifications the applicant acknowledged a need for a limitation, but also indicated a continuing need for improvement in the statement of the surveil-lance requirement. We had discussions with the applicant's staff on this subject. We will continue our review of the applicant's proposed improvement.

s - -- -- .-.-, .. - _

3.3.1 As a minimum, the reactor trip systen instrumentati,on, channels v and interlocks of Table 3.3-1 shall be OPERABLE [with RESPDNSE TIMES as shown in Table 3.3-2].

a 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumen-tation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint Column of Table 3.3-4 [and with RESPONSE TIMES as shown in l

Table 3.3-5).

The staff position is that the elimination of RESPONSE TIMES from the LCOs of 3.3.1 and 3.3.2 is acceptable as a technical specifi-cation simplification. Because the definition of OPERABLE includes response time there is no relaxation of requirements by this change.

However, the RESPONSE TIMES that were provided in the Technical Specifications should be transferred to the FSAR. During the Tech-nical Specification review the staff will ensure that this transfer j_ has occurred.

• t i

C. Deletion of Surveillance Requirements i

4.4.1.1 The above required reactor coolant loops shall be verified

! to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

l The applicant argues that Surveillance 4.4.1.1 is unnecessary because

! the operator is constantly observing his control boards which contain alarms, indications, and graphs. The staff position is that the I

surveillance must be retained because the flow rate is a fundamental parameter important to assuring that operation is within safety i

limits, and hence the cost of recording the verification once each 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is small compared to the value of assuring operation within safety limits.

i

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< 4.4.9.1.2 The reactor vessel material irradiation surveillance speci-mens shall be removed and examined, to determine changes is material properties as required by 10 CFR 50, Appendix H, in accordance with the schedule in Table 4.4-5. The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3.

Surveillance 4.4.9.1.2 does not include any requirements not already incorporated in 10 CFR 50, Appendix H. Therefore, the staff position is that Surveillance 4.4.9.1.2 can be deleted.

4.7.7 (all of the surveillance - 5 pages of text and 1 figure)

The applicant proposes to delete Surveillance Requirements 4.7.7 for snubbers and to include the requirements in a licensee main-tained and controlled document. The Bases state "All snubbers are required to be OPERA 8LE to insure that the structural integrity of the Reactor Coolant System and all other safety-related systems is maintained during and following a seismic or other event initiating dynamic loads." Clearly, snubbers arp active devices that may have to function upon demand to avoid an immediate degradation of capability required for mitigating the consequences of design basis events.

Therefore, the staff position is that the Surveillance Requirement should be retained within the Seabrook Technical Specifications.

However, the staff agrees that if an inservice inspection program is described in the FSAR, and is reviewed and approved by the staff, 1 the surveillance requirement in the Technical Specifications can l

be limited to this requirement

i "4.7.7 Each snubber shall be demonstrated OPERABLE by performance of an approved inservice inspection program".

e, , - - - , , -

D. Deletten of Tables 3.2-1 DNB Parameters, .

y 4.3-5 Meteorological Monitoring Instrumentation Surveillance Requirements, a

4.3-6 Remote Shutdown Monitoring Instrumentation Surveillance Requirements, i

4.3-7 Accident Monitoring Instrumentation Surveillance Requirements, 4.4-3 Reactor Coolant System Chemistry Limits Surveillance Requirements With few exceptions the material in these tables is a repetition <

of material in other tables. For the exceptions, simple text

• additions can be used in lieu of the tables. Thus, these deletions are in the nature of editorial simplifications and are acceptable.

3.7-4 Fire Hose Stations 3.7-5 Yard Fire Hydrants and Hydrant Hose Houses

' The ACTIONS associated with these tables do not directly limit operation of the facility. Therefore, maintenance of the 4 tabulations in the FSAR instead of in the Technical Specifications is appropriate and the tables can be deleted.

3.6-2 Containment Isolation Valves 3.8-1 Containment Penetration Conductor Overcurrent

• Protective Devices 3.8-2 Motor Operated Valves Thermal Overload Protection Devices Because the surveillance requirements will be retained these tables can be deleted from the Technical Specifications and the information included in the FSAR. The provisions of 10 CFR 50.59 will provide adequate opportunity for review of changes to this information by the staff.

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. - - - . - - - ,-r.- . . - _ - - - , - . - < _ __m--__.-~._ _ ___ - - - - - - - - - - - -- - - . . - -

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

.g. l i

4.-11-1 Radioactive Liquid Waste Sampling and Analysis Program, 4.11-2 Radioactive Gaseous Waste Sampling and Analysis Program, s

3.12-1 Radiological Environmental Monitoring Program, 3.12-2 Reporting Levels for Radioactivity Concentrations in Environmental Samples, and 4.12-1 Detection Capabilities for Environmental Sample Analysis These tables can be included in the off-site Dose Calculation Manual (00CM) instead of in the FSAR. To preclude unacceptable changes, Commission approval will be required prior to implementing changes to these tables.

0 E. Extension of Surveillance Intervals 9

4.5.1.1.1 Each accumulator shall be demonstrated

't OPERABLE:

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by:

1. Verifying, by the absence of alarms, the contained berated water volume and nitrogen cover pressure in the tanks, and i

2. Verifying that each accumulator isolation valve is open.

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The staff position is that the extension of the surveillance interval to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (from 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) is acceptable. .The staff has reviewed the applicant's probabilistic assessment and concurs that the contribytion to risk from extension of this surveillance interval is small compared to other potential failures in the ac-j cumulator trains.

4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that the following valves are in the indicated positions with power to the valve i operators removed:

The staff position is that the extension of the surveillance interval to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (from 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) is acceptable. The staff has reviewed the applicant's probabilistic assessment and concurs that the contribution to risk from extension of this surveillance interval is small compared to other potential failures in the ECCS subsystems.

, F. Extension of Action Time Limit 2.1.1 Safety Limits - Reactor Core l

2.1.2 Safety Limits - RCS Pressure l

l 2 The staff position is that the one-hour action time for reactor l

, trip if the safety limits are violated be retained. The basis is 1

that violation of a safety limit without an automatic plant trip indicates an unreviewed gap or failure in the plant protection system. Such an event is a serious safety concern. The reactor should be tripped expeditiously and a thorough review performed to understand the root cause of the failure.

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-- ,- , ~,. -

n . ,-, - - , - - - - , . - , - _,--_.-.-,,--.---..-,.-,..,-...--,c.- -

.-------,-.-----n_.-

8-1 G. Extension of Allowed Outane Times (A0T)

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j 3.5.1.la With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERA 8LE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT STAN08Y...

i The staff position is that the extension of this ACT to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (from I hour) is acceptable. The basis is that the contribution to risk from this A0T is small and the necessity that all four accumulators be operable is conservative for risk assessment.

3.5.2 With one ECCS subsystem inoperable, restore the inoperable i

subsystem to OPERABLE status within 7 days or be in at least HOT STAND 8Y...

f The staff position is that the extension of this A0T to 7 days (from 3 days) is acceptable. The staff has reviewed the applicant's probabilistic assessment of this change in A0T and concurs that the

! contribution to risk is small. The relaxation of this ACT does not i

relieve the applicant's responsibility for performing repairs to the

, system in a timely manner.

1

! H. Reduction of Allowed Outane Time 3.8.1.1.a/b

... restore at least two offsite circuits to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and two diesel generators to 0PERA8LE status within 72 I

hours...

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• The staff position is that the A0T for an offsite circuit be reduced to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />). The applicant proposed an extension of the A0T to 7 days based on his probabilistic assessment of station blackout events. The staff has, reviewed the probabilistic assessment and supporting data on the failures and repair times for the two SF6 lines into the Seabrook Station.

We note that the failure rates for the SF6 lines are high and the repair times are longer than those observed for aerial lines.

{ The applicant claims these data are non-representative because the plant is under construction and the SF6 lines are not receiving close attention. A published article that surveyed SF6 installa-tions (cited by the applicant) indicates a four year break-in period i and substantial repair times for certain types of failures. In the staff's opinion, the available evidence suggests that the SF6 lines at Seabrook will have poorer performance than aerial line installa-j tions in the near term. A revised probabilistic assessment using

- SF6 line data (from Seabrook) supports a reduction in the A0T for the SF6 to reduce the contribution of Seabrook Station blackout events to risk. The applicant may want to reexamine this A0T if he increases the number of offsite a c lines into the site or develops a history of favorable experience.

)

l I. Other 3.7.1.1.c With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater l

t pump to OPERA 8LE status as soon as possible.

The staff position is that this action statement for the auxiliary feedwater pumps be retained. The staff agrees that a precipitous plant trip should not be implemented because AFW is normally used

! for decay heat removal in the shutdown mode. The staff does not ,

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interpret the standard technical specification to require a reactor trip when there is no AFW available. The situation posed by the loss of AFW should be covered by emergency procedures that define alternate actions for decay heat removal (such as use,of the main feedwater system) if the reactor trips and other prudent steps such as power reduction. Power reduction should be performed slowly and maintained within a range in which the plant is known to be stable.

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Enclosure 2 -

Seabrook Technical Specification Improvement Procedures The applicant proposed to transfer items removed from the Seabrook Station

- Technical Specifications to the Nuclear Production Requirements' Manual.

The staff's position is that such information should be included as an appendix to Chapter 16 of the FSAR and thus avoid the introduction of another document for staff review. The remainder of the Nuclear Production Requirements Manual that delineates the procedures to be used should also be included by reference in the FSAR and should be submitted for staff review, and should be referenced in Section 6.8.1 of the Technical Specifications as a manual for an activity for which written procedures shall be established, implemented and maintained.

The applicant proposed to submit a separate report after each individual 4 change made under 10 CFR 50.59. The staff's position is that submission

! of the summaries of these reports at regular intervals as provided by 10

) .

CFR 50.59 is more appropriate, and the staff is considering a license con-i dition that would establish that interval as quarterly rather than annually i which is pemitted by 10 CFR 50.59. This woulti provide early opportunity

. for staff review while allowing for regularly scheduled report submittals under 10 CFR 50.59.

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Our basic deci.sion to approve transfer of items from the Seabrook Technical Specifications to an'appendjx to Chapter 16 of the Seabrook FSAR was made with the thought that the provisions of 10 CFR 50.59 would be applicable to subsequent changes to that infomation. However, the three categories of changes for which the provisions of 10 CFR 50.59 are applicable do not clearly include all of the changes we proposed to approve, e.g., unit staff qualifications. We are considering the development of an appropriate license condition to cover such changes.

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1 4 Enclosure 7-A. .-

Westinghouse Water Reactor amm Electric Corporation Divisions PmmePsevenin NSNRC-84-2990 1

December 21, 1984 l

Dr. Cecil 0. Thomas, Branch Chief Standardization and Special Projects Branch Office of Nuclear Reactor Regulation U.S. Nuclear Regulatcry Comission Washington, D.C. 20555

Dear Dr. Thomas:

i

! Attached please find a Westinghouse recomended revision to Specification 1

3/4.1.3 Movable Control Aasemblies and its associated Bases. This revision is the result of efforts made to address multiple imovable, but trippable,

!- control reds. These efforts were made due to several occurrences at different plants with a Westinghouse KSSS where a group or several groups of control rods i

became imovable by a md control system failure. Under these conditions the.

control rods would not step in or out, but would drop if a reactor trip was initiated. ,

' The Technical Specifications as currently provided in NURED-0452 RE.V. 4 do not j

recognize the fact that in this situation the control rods would still perfonn their safety flaiction and becatne more than one control red is imovable the plar.t is forced to repair the failure or be in Hot Shutdown in six hours. An action this drastic is unnecessary as noted by Westinghouse verbally in recent discussions with members of the staff.

It should be noted that the suggested revisions have been discussed in detail I

with Mr. M. S. Dunenfeld of the Core Perronnance Branch. It is our i

understanding that he is in general agreenent with the proposed changes based upon our subnittal of additional infonnation concerning rod control systen

, organization, failure, and troubleshooting. The requested infonnation is j provided as Attachnent C.

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e Dr. C. O. Th m.as i

Page h o It is our belief that the proposed changes to this specification adequately address the safety requirenents with regards to imovable and misaligned control rods. It would be appreciated if you would review the attached and consider its inclusion in all Westinghouse NSSS plant Technical Specifications (if so requested by the plants) and any future revisions to NURED-0452. If you have any questions concerning the above or the attached please contact either Mr. W. L. Luce, Manager, Licensing Initiatives (412/374-4793) or Mr. C. R.

Tuley (412/374-4172).

Very truly yours,

• h u> btw W P. Rahe, Jr., Manager uclear Safety Department CRT/kk Attactinents cc: D. S. Brinkman M. S. Dunenfeld 4

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ATTACMENT A REAETIVITY CONTROL SYSTms 4

~ "4/4.1."4 MOVAR1 F CONTROL ARMFMBL TFR Git 00P MEIGHT j LIMITTNG CONDITION FOR OPERATION 3.1 3.1 All shutdown and control rods, which are inserted in the core, shall be OPERABLE and positioned within i 12 steps (indicated position) of their group step counter d eand position.

APPLfrARTITTY. Hgpg3 je AND 2*.

EIIQ1

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a. With one or more rods inoperable due to being imovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 31.1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

~

b. With one rod trippable but inoperable due to causes other than j

addressed by ACTION a above, or misaligned from its group step counter j deand height by more than A 12 steps (indicated position), POWER 1

OPERATION may continue provided that within one hour either:

1. The rod is restored to OPERABLE status within the above alignment requirements, or

2. The md is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within i 12 steps of the inoperable md while maintaining the rod sequence i and insertion limits of Figure 3.1-1. The THERMAL POWER level shall be restricted pursuant to Specification 313.6 during

subsequent operation, or

3. The rod is declared inoperable and the SHUTDOWN MARGIN j requirement of Specification 31.1.1 is satisfied. POWER j OPERATION may then continue provided that:

! a) A reevaluation of each accident analysis of Table 31-1 is performed within 5 days; this reevaluation shall confiru i that the previously analyzed results of these accidents remain valid for the duration of operation tmder these

! conditions.

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b) A power distribution mapJs obtained from the movable incore '

detectors limits withinand 72Fo(Z) and rih are verified to be within their hours.

l

• S,.e Special Test Exceptions 310.2 and 310 3 l

, . _ - - - . - - - . - - - - - - - - - - - - - - - - - - - ' ' ~ ~ ~ ~ ~ ~ ^~

1

• c) The THERMAL POWER level is reduced to less than or equal to 755 of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the high neutron flux trip setpoint is reduced to less than or equal to 855 of RATED THERMAL PU.G.:..

c. With more than one rod trippable but inoperable due to causes other than addressed by ACTION a above, POWER OPERATION may continue

, provided that:

1.

Within one hour, the remainder of the rods in the bank (s) with the inoperable rods are aligned to within A 12 steps of the

- inoperable rods while maintaining the rod sequence and insertion limits of Figure 3.1-1. The THERMAL POWER level shall be restricted pursuant operation, and to Specification 3.13.6 during subsequent

'm

2. The inoperable rods are restored to OPERABLE status within 72

hours.  !

O

d. With more than one rod misaligned from it's group step counter demand height by more than A 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SJRVFTt f ANCE RrotJTRDENTE 4.1 3 1.1 The position of each rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the Rod Position Deviation Monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4.1 3.1.2 Each red not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days.

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TABLE 31-1 ACCTBDJT ANAf YRTR RF3UIRTNO RTWALUATION .

IN THE EVDJT OF AN TNOPERART F ROD Rod Cluster Control Assembly Insertion Characteristics Fod Cluster Control Assembly Misalignment Less of Reactor Coolant From Small Ruptured Pipes or From Cracks In Large Pipes Which Actuates the Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal At Full Power Major Reactor Cociant System Pipe Ruptures (Loss of Coolant Accident)

Major Secondary System Pipe Rupture Rupture Of A Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejectio.1) 4

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

i ATTACIMENT B M

1 / 41 . 1 . 1 Mavahl e Centrol Amesmh14em

^:

The specifications of this section are necessary to ensure that the following requirements are met at all times during normal N loop or N-1 loop operation.

I Sy observing that the RCCAs are positioned above their respective insertion .,,

l limits during normal operation,

1. At any time in life for Mode 1 and 2 operation, the minimm SHUTDOWN MARGIN will be maintained. For operational modes 3, 4, 5, and 6, the reactivity 4

condition consistent with other specifications will be maintained with all RCCAs fully inserted by observing that the boron concentration is always greater than an appropriate minizia value. ,

2. During normal operation the enthalpy rise hot channel factor, F will be i

maintained within acceptable limits. 4He 3 The consequences of an ejected RCCA accident will be restricted below the limiting consequences referred to in the ejected rod analysis.

j 4. The core can be made suberitical by the required shutdown margin with one j RCCA stuck. In the event of an RCCA ejection, the core can be made subcritical with two RCCAs stuck, where one of the RCCAs is asstmed to be j the worst ejected rod control assembly. - "-

5. The trip reactivity assimed in the accident analysis will be available.

6. Dropping an RCCA into the core or statically misaligning an RCCA during  ;

i nomal operation will not violate the themal design basis witn respect to l DNBA.

I 7. The uncontrolled withdrawal of an RCCA will result in consequences no more severe than presented in the accident analysis.

I 8. The econtrolled withdrawal of a control assembly bank will not result in a

peak power density that exceeds the center line melting criterion.

OPERABILITY of the control rod position indicator channels (LCo 31.3 2) is '

required to detemine control rod positions and thereby ensure compliance with

the control rod alignment.

OPERABILITY of the demand position indication system (LCO 313 2) is required t

to detemine bank demand positions and thereby ensure compliance with the

! insertion limits.

i

The ACTION statements which permit limited variatior.s from the basic

! requirenents are accompanied by additional restrictions which ensure that some l of the original criteria are met. Misalignment of a rod requires measurement

! of peaking factors or a restriction in THERMAL POWER, either of these restrictions provide assurance of fuel rod integrity during continued operation provided no further abnormal condition develops.

i Tcr Specification 3131 ACTIONS b and e it is incumbent upon the plant to verify the trippability of the inoperable control rod (s). This may be by ~

i verification of a control system failure, usually electrical in nature, or that l the failure is associated with the control red stepping mechanism. In the t

i

\

event tae plant is unable to verify the rod (s) trippability, it must be asstraed '-

to be mtrippable and thus fall saider the requirements of ACTION A. Asstating a

, controlled shutdown from 1005 RATED THERMAL POWER, this allows apprtximately four hours for this verification.

The maxista rod drop time permitted by (LCO 313 4) is consistent with the assumed rod drop time used in the accident analyses. Measurement with 7 350 degrees-F and with all reactor coolant pumps operating ensures that* Die measured drep times will be representative of insertion times experienced during a reactor trip at operating conditions.

Bank demand positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCO's are satisfied.

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ATTACHENT C ,

i Arrangement of Mechani ms The Westinghouse NSSS design for four, three, and two loop plants uses fitty-three, forty-eight and twenty-nine full length rod drive mechanies j arranged into banks and groups as shown in Figure 1. A group consists of two or i more mechanisms that are electrically paralleled to step simultaneously. A bank i of mechanisms consists of two groups that are awed in staggered fashion such j that the groups are always within one step of each other. Part length rods have i either been remwed or locked out in all Westin$ouse NSSS plants. The arrangement includes two or four shutdown banks, A and B or A, B, .C and D; and f

. four Control Banks A, B, C and D. Control banks are moved in worlap in the

! folicwing withdrawal sequence: When Control Bank A reaches a predetermined i height in the top half of the core, Control Bank B starts to see out with A.

. Control Bank A stops at the top of the core and Bank B continues until it i reaches a predetermined height in the top half of the core when Control Bank C

] starts to swe out with B. This sequence continues until all rods are j

withdrawn. The insertion sequence is the opposite of the withdrawal sequence.

Main Central llaan Contrala

Controls for the Rod Control Systen located in the main control room are listed 1 ' '

in Figure 2. The Ir> Hold-Out lever is a joy stick used for manual twd motion

and is located on the main contrcl board. Also on the main ocntrol board is a

Bank Selector .% itch with eight positions. In the manual position, control l banks are swed in overlap with the Ir> Hold-out lever. Control banks are awed

! in worlap by the automatic Tavg control system with the atitch in the auto 4

position. Six additional positions are provided for individual bank sweent.

j Step counters, one for each group, are located on the main ocntrol board to dispisy demand rod position. A Rod Position Indication System, not connected to

the Rod Control Statem, is used to dispisy actual red position and is used in I conjunction with the step counters to determine deviation between demand and ,

actual position.

1 Ir>0ut lights show the request for rod motion from either the In-Hold-out lever i or the Automatic Tavs Control Syste. A startup pushbutton is provided to reset i the step counters and all internal system counters such as the bank overlap i counter on startup. An alarm reset pushbutton, resets internal system failure j detectors and alarms which include a seal in feature. Lift coil disconnect i switches, one for each mechanian, are provided behind the main control board to /

j allow retrieval of a dropped rod.

1Wo annunciators are located on the control board; an urgent and nor>. urgent j alarm. The urgent alarm indicates that a control system failure has occurred ,

! that would affect the ability of the control system to swe rods. A non-urgent j alarm indicates failures of one or more redundant power supplies that feed the f l system printed circuit cards.

j r i'

! s i Basic Thyrister Bridge Control Cir'cuit 2ree thyristors forming a half wave phase controlled bridge supply current to four mechanism coils, either lift, movable or stationary gripper coils as shown 4

in Figure 3. Current feedback signals from shtets in series with each coil are

, used to regulate the current comanded by a alave cycler located in the syste.

logic cabinet.

j l

Peer Cabinet Pser Circuits Five thyristor bridges fom one system power cabinet as shown in Figure 4.

Bree, four, or five power cabinets are used in the system. The power cabinet j basically unplifies low level comand signals from a slave cycler in the logie cabinet. One power cabinet drives three groups of four mechanians and is capable of moving one group while holding the other two in position. The selection as to which group is to move is made with multiplexing thyristors, one for each group of movable coils and one for each lif t coil. The lif t coil -

' multiplexing thyristors also serve as lif t disconnect switches for retrieving a dropped rod. ~

0 l Syntam Block Diagram

! he power cabinets are supplied with power from two motor generator sets j nomally operating in parallel through two reactor trip breakers in series as j shown in Figure 5. The logic cabinet includes a pulser, master cycler, bank

~ overlap tait, and four slave cyclers. The pulser determines the speed of red i motion as directed by the reactor Tavg control system when automatic operation i is selected or by a potentometer located in the Tavg control system when manual operation is selected. The master cycler directs pulses from the pulser

alternately to the slave cyclers for the two groups in a bank. Selection of I

which bank or banks are to move is made by the bank overlap tmit and master l -

cycler. De slave cycler sequences the mechanian coils through one step, either

! in or out for each ago" pulse from the master cycler. A DC hold cabinet is

} provided to hold rods and allow replacement of printed circuit cards in the 4

power cabinet while the plant is in operation.

Failure Detectim and 11ams

! A rod control urgent alarm is actuated by five failure detectors in each power l cabinet or by three failure detectors in the logic as shcun in Figure 6. An

urgent alam stops automatic rod motion and pemits manual movenent of a selected bank if the logic cabinet and the two power cabinets associated with the selected bank are not in urgent alarm.

l Detection of a failure by a failure detector results in the follcWing i indicaticms:

l

s. A failure detector lanp, one for each type of failure, located on the edge of a printed circuit card in the failed cabinet, is energized.  ;

j b. A red urgent failure lamp on the front of the failed cabinet is energized.

4

c. A ' Rod control Urgent Failure" annuncistor in the control roan is actuated. ,

i A non-urgent alarm indicates failure of one of a runber of redundant power j supplies and does not affect the operation of the system.

l 4

. Effret of Mechanism Mechanical Failu:t on Contdl System Th2 Rod Control System operat:s independ:ntly of tha rod drive m:chanims.

Nothing has been included in the system, either by design or iradvertently, to

• allow it to see movement of the medianiam mechanical parts, the drive shaft, or rod contrcl clusters. This has been verified many times during factory check out of completed systens where the test loads consist only of simulated

~

mechanim coils with iron pipes in the center to approximate the magnetic properties of the coils. Q)eckout of the Rod Control system at the site is generally perfomed during the Hot Functional Test when no reactor core, rods, or drive shafts are in place but only the mechanim coils are connected. In the factory tests and Hot Functional checkout, the systems operate nomally without alams.

Hm to Distinnuish Between Control Syste and Mechanim Preblarna Based on the previous discussion, a Rod Control Systen Urgent Alam mast be the ,

result of a control system failure and cannot be related to an inoperable rod or rods. There are failures that do not result in an urgent alam that could prevent one or more rods from moving. In this case, the problen can be traced to either the control system or mechania by monitoring the mechanim coil i

currents with a voltmeter, oscilloscope, or recorder. Built-in test points are located in the power cabinets for this purpose. If the control system will not vary the currents to the mechanisn coils, the problem Eat be in the control system and not the mechanims. If the control system varies currents to the coils, then the mechanim may be suspect. Grossly abnomal currents would indicate control system problens and mildly abnomal currents would indicate mechanian problens. Recordings of the currents would have to be studied in this

_ event. Figure 7 stamarizes this identification process.

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Total Number of Mechanisms el 53 33 Full t.ength Mechanisms 53 48 l 29 w . . .bi '

' Figure 1.

Typical Irrangements of Mechanium Banks and Groups l

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MAIN CONTROL ROOM CONTROLS o IN-HOLD-0UT LEVEL o BANK SELECTOR SWITCH MANUAL

, AUTO SHUTDOWN BANK A, B 4

CONTROL BANK A, B, C, D o STEP COUNTERS SHOW DEMAND POSITION i ' o IN-0UT LIGHTS l

c STARTUP PUSHBUTTON o ALARM RESET PUSHBUTTON ,, .,. -

o LIFT COIL DISCONNECT SWITCHES

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o ANNUNCIATORS ROD CONTROL URGENT ALARM ROD CONTROL NON-URGENT ALARM FIGURE 2 l

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URGENT ALARM - POWER CABINET -

o REGULATION FAILURE o PHASE FAILURE o LOGIC ERROR o MULTIPLEXING ERROR o CARD MISSING URGENT ALAR $ - LOGIC CABINET o

SLAVE CYCLER RECEIVES A GO PULSE DURING A STEP.

o OSCILLATOR FAILURE o CARD MISSING URGENT ALARM EFFECT ON SYSTEM l o AUTOMATIC ROD MOVEMENT IS STOPPED o

MANUAL MOVEMENT OF SELECTED BANK IS PERMITTED IF lLOGIC CABINET AND POWER CABINET ARE NOT IN URGENT ALARM NON-URGENT ALARM - LOGIC OR POWER CABINET i

o REDUNDANT POWER SUPPLY FAILURE FIGURE 6

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HOW TO DISTINGUISH BETWEEN CONTROL SYSTEM AND MECHANISM PROBLEMS o URGENT ALARM - MUST BE CONTROL SYSTEM FAILURE o NO URGENT ALARM, ONE OR MORE RODS WON'T MOVE MONITOR COIL CURRENTS CONTROL SYSTEM WON'T VARY CURRENTS +

PROBLEM MUST BE IN CONTROL SYSTEM CONTROL SYSTEM VARIES CURRENT 70 C0ILS NORMALLY + SUSPECT MECHANISM CURRENTS ABNORMAL GROSSLY ABNORMAL + SUSPECT CONTROL SYSTEM MILDLY ABNORMAL + SUSPECT MECHANISM i

FIGURE 7

Enclosure 8 CONTAINMENT SYSTEMS CONTAINMENT VENTILATION SYSTEM j

j LIMITING CONDITION FOR OPERATION 3.6.1.7 Each containment purge supply and exhaust isolation valve shall be OPERABLE and:

l

a. Each 42-inch containment purge supply and exhaust isolation valve shall be sealed closed.

b. The 8-inch containment purge supply and exhaust isolation valves l shall be sealed closed to the maximum extent practicable but may be open for purge system operation for pressure control, for ALARA and respirable air quality considerations for personnel entry and for ,

surveillance tests that require the valve to be open.  !

APPLICABILITY: MODES 1, 2, 3, and 4.  ;

ACTION:

a. With a 42-inch containment purge supply and/or exhaust isolation valveb) open or not sealed closed, close and/or seal close the open valve (s) or isolate the penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, otherwise be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With an 8-inch containment purge supply and/or exhaust isolation valve (s) open for reasons other than given in 3.6.1.7.b above, close the open 8-irch valve (s) or isolate the penetration (s) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, ,

' otherwise be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in - ,

COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With a containment purge supply and/or, exhaust isolation valve (s) having a measured leakage Tate exceeding the limits of Specifica-tions 4.6.1.7.2 and/or 4.6.1.7.3, restore the inoperable valve (s) to OPERABLE status or isolate the penetrations such that the measured leakage rate does not exceed the limits of Specifications 4.6.1.7.2 and/or 4.6.1.7.3 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, otherwise be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN j within'the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

j SURVEILLANCE REQUIREME! S 4.6.1.7.1 Each 42-inch containment purge supply and exhaust isolation valve shall be verified to be sealed closed at least once per 31 days.

4.6.1.7.2 At least once per 6 months on a STAGGERED TEST BASIS each sealed closed 42-inch containment purge supply and exhaust isolation valve with resilient material seals shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than or equal to 0.05 L, when pressurized to P,.

4.6.1.7.3 At least once per 92 days each 8-inch containment purge supply and exhaust isolation valve with resilient material seals shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than or equal to 0.01 L, when pressurized to P,.

4.6.1.7.4 Each 8-inch containment purge supply and exhaust isolation valve I l

shall be verified to be sealed closed or open in accordance with specifica-tion 3.6.1.7.b at least once per 31 days.

PALO VERDE - UNIT 2 3/4 6-14 1

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UNITED 3TATEs nclosure 10 I; g NUCLEAR REGULATORY COMMISSION a ;; . WASHINGTON, D. C. 20SSS p

1. . '*,'.'..p' June 22, 1984 Docket Nos. 50-348 and 50-364 i

Mr. R. P. Mcdonald a Senior Vice President I Alabama Power Company Post Office Box 2641 Binningham, Alabama 35291 ,

Dear Mr. Mcdonald:

The Connission has issued the enclosed Amendment No.46 to Facility

- Operating License No. NPF-2 and Amendment No.37 to NPF-8 for the Joseph M. Farley Nuclear Plant, Unit Nos.1 and 2, respectively. The amendments consist of changes to the Technical Specifications in response to your application transmitted by letter dated March 4,1983, supplemented March 1, 1984 The amendments modify Technical Specifications to clarify and update the charcoal filter surveillance testing. 5pecific filter test efficiency

- requirements are shown rather than referencing Regulatory Guide 1.52. An existing error is also corrected since.the plant design does not have a

( bypass filter system.

A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Comission's next regular monthly Federal Register notice.

i Sincerely, Q vt G: W Edward A. Reeves, Project Manager Operating Reactors Branch #1 Division of Licensing

Enclosures:

1. Amendment No.46 to NPF-2

2. Amendment No.37 to NPF-8

3. Safety Evaluation cc: w/ enclosures See next page

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Mr. R. P. Mcdonald Joseph M. Farley Nuclear Plant Alabama Power Company Units 1 and 2 cc: Mr. W. O. Whitt D. Biard MacGuineas. Esquire Executive Vice President Volpe, Boskey and Lyons I

Alabama Power Company 918 16th Street, N.W.

Post Office Box 2641 Washington, DC 20006 l Birmingham, Alabama 35291 Charles R. Lowman Mr. Louis B. Long, General Manager Alabama Electric Corporatien Southern Company Services, Inc. Post Office Box 550 Post Office Box 2625 Andalusia, Alabama 36420 Birmingham, Alabama 35202 James P. O'Reilly Houston County Commission Regional Administrator - Region II Dothan, Alabama 36301 U.S. Nuclear Regulatory Commission 101 Marietta Street, Suite 2900 Robert A. Buettner, Esquire Atlanta, GA 30303 George F. Trowbridge, Esquire Shaw, Pittman, Potts and Trowbridge 1800 M Street, N.W.

Washington, DC 20036

. Chai rman I

- _ Houston County Conmission Dothan, Alabama 36301

(

Robert A. Buettner, Esquire

' Balch, Bingham, Baker, Hawthorne, Williams and Ward Post Office Box 306 Birmingham, Alabama 35201 a

Resident inspector U.S. Nuclear. Regulatory Commission Post Office Box 24 - Rnute 2 Columbia, Alabama 36319 State Department of Public Health ATTN: State Health Officer State Office Building Montgomery, Alabama 36104 l Regional Radiation Representative EPA Region IV 345 Courtland Street, N.E.

Atlanta, GA 30308 .

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[R,$*'*'gtug,#g, UNITED STATES .

NUCLEAR REGULATORY COMMISSION 7, . ' j WASHINGTCN. D. C. 20585

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ALABAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 46 License No. NPF-2

1. The Nuclear Regulatory Connission (the Connission) has found that:

A. The application for amendment by Alabama Power Company (the licensee) dated March 4, 1983, supplemented March 1, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Connission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate inIconfortnity with the application, as amended, the provisions of the Act, and the regulations of the Commission; .

C. There is reasonable assurance: (1) that the activities authorized by this amendment can be conducted without endangering the health and ,

safety of the public, and (ii) that such activities will be conducted in compliance with the Connission's regulations; D. The issuance of this license amendment will not be inimical to the connon defense and security or to. the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51

, of the Connission's regulations :nd all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating. License No. NPF-2 is hereby amended to read as follows:

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(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.46 , a re  !

hereby incorporated in the license. The licensee l shall operate the facility in accordance with the '

Technical Specifications,

3. This license amendment is effective as of its date of issuance.

t FOR THE NUCLEAR REGULATORY COPHISSION Operating Reactors nch #1 Division of Licensin

Attachment:

Changes to the Technical .

Specifications -T Date of Issuance: June 22, 1984 . .

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ATTACHMENT TO LICENSE AMENDMENT NO. 46 i AMENDMENT NO.46 FACILITY OPERATING LICENSE NO. NPF DOCKET NO. 50-348 Revised Appendix A as follows: j l

Insert Pages Remove Pages 3/4 7-16 3/4 7-16 3/4 7-17 3/4 7-17 3/4 7-17a 3/4 7-17a 3/4 7-18 3/4 7-18 3/4 7-19 3/4 7-19 3/4 9-17 3/4 9-17 3/4 9-18 3/4 9-18 t -

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o PLANT SYSTEMS

, 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.7 Two independent control room emergency air cleanup systems shall be OPERABLE.

APPLICABILITY: ALL MODES.

ACTION:

MODES '1, 2, 3 and 4:

With one control room emergency air cleanup system inoperable, restore the inoperable system to OPERA 8LE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

MODES 5 and 6 (during irradiated fuel movement, or movement of loads over irradiated fuel):

a. With one control room emergency air cleanup system inoperable, restore the inoperable system to OPERABLE status within 7 days

• or initiate and maintain operation of the control room emergency ventilation system in the recirculation mode.

b. With both control room emergency air cleanup systent, inoperable, suspend all operations involving the movement of irradiated fuel or movement of loads over irradiated fuel.

c. The provisions of Specification 3.0.3 are not applicable in MODE 6.

SURVEILLANCE REQUIREMENTS 4.7.7 Each control room emergency ventilation system shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the-control room air tagerature is less than or equal to 120*F.

b. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the pressurization and recirculation system HEPA filters and charcoal adsorbers and verifying that the system has operated for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters on during the past 31 days.

c. At least once per 18 months or (1).after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release that could I have contaminated the charcoal adsorbers or HEPA filters in any ventilation zone communicating with the system by:

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FAALEY-UNIT 1 3/4 7-16 AMENDMENT NO.46

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PLANT SYSTEMS  !

, SURVEILLANCE REQUIREMENTS (Continued)

1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria of greater than or equal to 99.5% filter efficiency while operating the system at a ,

flow rate indicated in Note 1 and using the following test I procedures:

1 (a) A visual inspection of the control room emergency air cleanup system shall be made before each DOP test or activated carbon adsorber section leak test in accordance with Section 5 of ANSI N510-1980.

(b) An in-place DOP test for the HEPA filters shall be performed in accordance with, Section 10 of ANSI N510-1980.

(c) A charcoal adsorber section leak test with a gaseous halogenated hydrocarbon refrigerant shall be performed in accordance with Section 12 of ANSI N510-1980.

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in

- accordance with Section ITof ANSI N510-1980 meets the laboratory testing efficiencies criteria given in Note 2

when tested with methyl iodide at 80*C and 70% relative

( humidity.

3. Verifying a system flow rate as indicated in Note 1 during system operation when tested in accordance with Section 8

, of ANSI N510-1980.

d. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by i verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Section 13 of ANSI N510-1980 meets the i laboratory testing efficiencies criteria given in Note 2 when tested with methyl iodide at 80*C and 70% relative humidity.

e. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the system at a flow rate indicated in Note 1.

2. Verifying that the filter train starts on a Safety Injection Actuation test signal.#

1. Surveillance Requirement 4.7.7.e.2 dces not apply in MODES 5 and 6.

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FAkEY-UNIT 1 3/4 7-17 AMENDMENT NO. 40 xA t

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

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

3. Verifying that the system maintains the control room at a positive pressure of greater than or equal to 1/8 inch water guage relative to the outside atmosphere during system operation.

4. Verifying that the heaters dissipate 7.5 + 0.8 kw when tested in accordance with Section 14 of ANSI N510-1980.

f. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.5% of the DOP when they are tested in-place in accordance with Section 10 of ANSI N510-1980 while operating the system at a flow rate indicated in Note 1.

g. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.5% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with Section 12 of ANSI N510-1980 while operating the system at a flow rate indicated in Note 1.

Note 1. a. Control Room Recirculati n Filter Unit 2000 cfm + 10%

, - b. Control Room Filter Unit 1000 cfm T 10%

( c. Control Room Pressurization Filter Unit 300cfmi10%

• Note 2. a. Control Room Recirculation Filter Unit > 99%

b. Control Room Filter Unit 7 99%

c. Control Room Pressurization [99.8255 l

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FARLEY-UNIT 1 3/4 7-17a AltEN0tiENT NO.46

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PLANT SYSTEMS

. 3/4.7.8 PENETRATION ROOM FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.8 Two independent penetration room filtration system's shall be OPERABLE.

I APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:

With one penetration room filtration system inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

i . SURVEILLANCE REQUIREMENTS 4.7.8 Each penetrat3on room filtration system shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control toom, the flow through the HEPA filters and charcoal adsorbers and verifying that the system has operated for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters on during i

( the past 31 days.

b. At least once per 18 months or (1) after any structural

- maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release that could have contaminated the charcoal adsorbers or HEPA filters in any

!,! ventilation zone comunicating with the system by:

1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria of greater than or equal to 99.5% filter efficiency while operating the system at a flow rate of 5000 cfm + 10 percent and using the following test procedures

l (a) A visual inspection of the penetration room filtration system shall be made before each DOP test or activated carbon adsorber section leak test in accordance with Section 5 cf ANSI N510-1980.

1 (b) An in-place 00P test for the HEPA filters shall be

. performed in accordance with Section 10 of ANSI N510-1980.

I (c) A charcoal adsorber section leak test with a gaseous halogenated hydrocarbon refrigerant shall be performed in accordance with Section 12 of ANSI N510-1980, hlEY-UNIT 1 3/a 7-18 AMEN 0 MENT NO.46 x

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,, . , - - - - ,n -------,--,.,,-.,n------- - , - - , - - - - - - - - - , - - - ---

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Section 13 of ESI N510-1980 meets the laboratory testing criterion of greater than or equal to 95% efficiency when tested with methyl iodide at 80*C and 70% relative humidity.

3. Verifying a system flow rate of 5000 cfm + 10% during system operation when tested in accordance with Section 8 of ANSI N510-1980.

c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Section 13 of ANSI N510-1990 meets the laboratory testing criterion of greater than or equal to 95%

efficiency when tested with methyl iodide at 80*C and 70%

relative humidity.

d. At least once per 18 months by.;

1. Ve'rifying that the pressu e drop across the combined HEPA I

filters and charcoal adsorber banks of less than 6 inches Water Gauge while operating the system at a flow rste of 5000 cfm 110%.

2. Verifying that the system starts on a Phase B Isolation l test signal. l l

3, Verifying that the heaters dissipate 25 + 2.5 kw when tested in accordance with Section 14 of INSI N510-1980.

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.5% of the DOP when they are tested in-place in accordance with Section 10 of ANSI N510-1980 while operating the system at a flow rate of 5000 cfm 110%.

f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal tc 99.5% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with Section 12 of ANSI M510-1980 while operating the system at a flow rate of 5000 cfm + 10%.

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k FAIL,EY-UNIT 1 3/4 7-19 A!1ENDHENT NO.46

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REFUELING OPERATIONS

, SURVE!LLANCE REQUIREMENTS (Continued)

. 1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria of greater than or equal to 99.5% filter efficiency while operating the main purge system and using the following test procedures:

(a) A visual inspection of the containment purge exhaust filter system shall be made before each DOP test or activated carbon adsorber section leak test in accordance with Section 5 of ANSI N510-1980.

(b) An in-place DOP test for the HEPA filters shall be performed in accordance with Section 10 of ANSI N510-1980.

(c) A charcoal adsorber section leak test with a gaseous halogenated hydrocarbon refrigerant shall be performed -

in accordance with Section 12 of ANSI N510-1980.

. 2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in

- accordance with Section 13 of ANSI N510-1980 meets the laboratory testing criterion of greater than or equal to 90% efficiency when tested with methyl iodide at 80*C and

( 70% relative hunidity.

b. After every 12 months of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Section 13 of ANSI N510-1980 meets the laboratory testing criterion of greater than or equal to 90%

efficiency when tested with methyl iodide at 80*C and 70s relative humidity.

c. At least once per 18 months by verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the main purge system.

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\ 3/4 9-17 AMENDMENT NO. 46 FAngEY-UNIT 1

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REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued)

d. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.5% of the DOP when they are tested in-place in accordance with Section 10 of ANSI N510-1980.

e. After each complete or partical replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.5% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with Section 12 of ANSI N510-1980 while operating the main purge system.

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3/4 9-18. AMEN 0 MENT NO. 46

\(ARLEY-UNIT 1

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UNITED STATES NUCLEAR REGULATORY COMMISSION io I

E wasMWGTON, D. C. 20555

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ALABAMA POWER COMPANY DOCKET NO. 50-364 JOSEPH M. FARLEY NUC'. EAR PLANT, UNIT NO. 2 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No.37 License No. NPF-8

1. The Nuclear Regulatory Comission (the Comission) has found that:

A. The application for amendment by Alabama Power Company (the licensee) dated March 4, 1983, supplemented March 1, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate inIconfomity with the application, as amended, the provisions of the Act, and the regulations of the Commission; -

(

C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and

- safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regi!1ations; D. The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-8 is hereby amended to read as follows:

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(2) Technical Specifications The Technical' Specifications contained in Appendices A and B. as revised through Amendment No.M , are hereby incorporated in the license. The Ticensee l shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COM11SSION

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rating Reactors B ch #1 Division of Licensing

Attachment:

Changes to the Technical 7 Specifications I Date of Issuance: June 22, 1984 i

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ATTACHMENT TO LICENSE AMENDMENT NOJ7 AMENDMENT NO 37 FACILITY OPERATING LICENSE NO. NPF-8 DOCKET NO. 50-364 Revised Appendix A as follows:

Remove Pages Insert Pages 3/4 7-16 3/4 7-16 3/4 7-17 3/4 7-17 3/4 7-17a 3/4 7-17a 3/4 7-18 3/4 7-18 3/4 7-19 3/4 7-19 3/4 9-17 3/4 9-17 3/4 9-18 3/4 9-18 i

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• PLANT SYSTEMS 3/4.7.7 CONTROL ROOM EERGENCY VENTILATION SYSTEM 1 lit 11 TING CONDITION FOR OPERATION 3.7.7 Two independent control room emergency air cleanup systems shall be OPERABLE.

APPLICABILITY: ALL MODES.

ACTION:

MODES 1, 2, 3 and 4:

With one control room emergency air cleanup system inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the

. following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

MODES 5 and 6 (during irradiated fuel movement, or movement of loads over irradiated fuel):

a. With one control room emergency air cleanup system inoperable, restore the inoperable system.to OPERABLE status within 7 days or initiate and maintain operition of the control room emergency ventilation system in the recirculation mode.

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k b. With both control room amergency air cleanup systems

inoperable, suspend all operations involving the movement o?

irradiated fuel or movement of loads over irradiated fuel.

c. The provisions of Specification 3.0.3 are not applicable in MODE 6.

SURVEILLANCE 'tEQUIREMENTS 4.7.7 Each control room emergency ventilation system shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the control room ~

air tangerature is less than or equal to 120*F.  ;

b. At least once per 31 days on a STAGGERED TEST BASIS by i initiating, from the control room, flow through the j pressurization and recirculation system HEPA filters and I charcoal adsorbers and verifying that the system has o'perated for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters on during the past 31 days.

c. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release that could

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have contaminated the charcoal adsorbers or HEPA filters in any g

ventilation zone communicating with the system by: l s

FAkEY-UNIT 2 3/4 7-16 AMEN 0 MENT NO. 37 s

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. 1 PLANT SYSTEMS

. SURVEILLANCE REQUIREMENTS (Continued) f

1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria of greater than or equal to 99.5% filter efficiency while operating the system at a flow rate indicated in Note 1 and using the following test procedures:

(a) A visual inspection of the control room emergency air cleanup system shall be made before each DOP test or activated carbon adsorber section leak test in accordance with Section 5 of ANSI N510-1980.

(b) An in-place DOP test for the HEPA filters shall be perfomed in accordance with Section 10 of ANSI N510-1980.

(c) A charcoal adsorber section leak test with a gaseous halogenated hydrocarbon refrigerant shall be performed -

in accordance with Section 12 of ANSI M510-1980.

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in

- accordance with Section 17 of ANSI N510-1980 meets the laboratory testing efficiencies criteria given in Note 2 when tested with methyl iodide at 80*C and 70% relative

( humidity.

3. Verifying a system flow rate as indicated in Note 1 during system operation when tested in accordance with Section 8 of ANSI N510-1980. ,

d. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Section 13 of ANSI N510-1980 meets the l laboratory testing efficiencies criteria given in Note 2 when l

tested with methyl iodide at 80*C and 70% relative humidity.

e. At least once per 18 months by:

! 1. Verifying that the pressure drop across the combined HEPA j filters and charcoal adsorber banks is less than 6 inches j Water Gauge while operating the system at a flow rate indicated in Note 1.

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2. Verifying that the filter train starts on a Safety i Injection Actuation test signal.#

! e Surveillance Requirement 4.7.7.e.2 does not apply in MODES 5 and 6.

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. F EY-UNIT 2 3/4 7-17 AltENDMENT NO.37 i \

PLANT SYSTEMS

, SURVEILLANCE REQUIREMENTS _(Contiinued)

3. Verifying that the system maintains the control room at a positive pressure of greater than or equal to 1/8 inch water gauge relative to the outside atmosphere during system operation.

4. Verifying that the heaters dissipate 7.5 + 0.8 kw when tested in accordance with Section 14 of AISI N510-1980.

f '. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.5% of the DOP when they are tested in-place i in accordance with Section 10 of ANSI N510-1980 while operating

• the system at a flow rate indicated in Note 1.  !

! g. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.5% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with Section 12 of ANSI M510-1980 while operating the system at a flow rate indicated in Note 1.

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_ Note 1. a. Control Room Recirculation Filter Unit 2000 cfm + 10%

b. Control Room Filter Unit 1000 cfm 7 10%

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c. Control Room Pressurization Filter Unit 300cfm310%

Neta 2. a. Control Room Recirculation Filter Unit > 99%

8"B . Control Room Filter Unit 7 99%

c. Control Room Pressurization 7

_ 99.825%

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3/4 7-17a AMEt4DMENT NO.37

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, PLANT SYSTEMS 3/4.7.8 PENETRATION ROOM FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.8 Two independent penetration room filtration systems sha*.1 be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:

With one penetration room filtration system inoperable, restors the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.8 Each penetration room filtration system shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by ,

initiating, from the control room, the flow through the HEPA -

filters and charcoal adsorbers and verifying that the system

'has operated for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters on during

( the past 31 days. .

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chmical release that could have contaminated the charcoal adsorbers or HEPA filters in any ventilation zone communicating with the system by:

I 1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria of greater or equal to 99.5%

filter efficiency while operating the system at a flow rate of 5000 cfm + 10 percent and using the following test procedures:

(a) A visual inspection of the penetration room filtration system shall be made before each DOP test or activated carbon adsorber section leak test in accordance with i Section 5 of ANSI N510-1900.

(b) An in-place 00P test for the HEPA filters shall be performed in accordance with Section 10 of ANSI N510-1980.

(c) A charcoal adsorber section leak test with a gaseous ,

halogenated hydrocarbon refrigerant shall be performed i in accordance with Section 12 of ANSI N510-1980. .

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3/4 7-18 AMEN 0 MENT NO.37 ( l FAkEY-UNIT~,

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$' PLANT SYSTEMS

. SURVEILLANCE REQUIREMENTS (Continued)

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Section 13 of ANSI N510-1980 meets the laboratory testing criterion of greater than or equal to

- 955 efficiency when tested with methyl iodide at 80*C and 70% relative hianidity.

3. Verifying a system flow rate of 5000 cfm + 10% during system operation when tested in accordance with Section 8 of ANSI M510-1980.

c. Af ter every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Section 13 of ANSI N510-1980 meets the

! laboratory testing criteria of greater than or equal to 95%

efficiency when tested with methyl iodide at 80*C and 70%

relative humidity.

d. At least once per 18 months by_:

1. Verifying that the pressub drop across the combined HEPA g

filters and charcoal adsorber banks of less than 6 inches Water Gauge while operating the system at a flow rate of

• 5000 cfm f,10%.

- 2. Verifying that the system starts on a Phase B Isolation test signal,

3. Verifying that the heaters dissipate 25 + 2.5 kw when tested in accordance with Section 14 of INSI N510-1980. .

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.5% of the DOP when they are tested in-place in accordance with Section 10 of ANSI N510-1980 while operating the system at a flow rate of 5000 cfm + 10%.

f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.5% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with Section 12 of ANSI N510-1980 while operating the system at a flow rate of 5000 cfm + 10%. ,

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F EY-UNIT 2 3/4 7-19 AMECMENT t40. 37 x

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{. REFUELING OPERAT20NS SURVEILLANCE REQUIREMENTS (Continued) 1 1

1. Verifying that the cleanup system satisfies the in-plac:t l testing acceptance criteria of greater than c,r equal to 99.5% filter efficiency while operating the raain purge system and using the following test procedures.

(a) A visual inspection of the containment purge exhaust filter system shall be made before each DOP test or activated carbon adsorber section leak test in '

accordance with Section 5 of ANSI N510-1980. l (b) An in-place 00P test for the HEPA filters shall be performed in accordance with Section 10 of ANSI N510-1980.

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(c) A charcoal adsorber section leak test with a gaseous halogenated hydrocarbon refrigerant shall be performed in accordance with Section 12 of ANSI N510-1980..

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in '

accordance with Section 13-of ANSI N510-1980 meets the laboratory testing criterfon of greater than or equal *4 90% efficiency when tested with methyl iodide at 80*C and 7

( 70% relative humidity.

b. Af ter every 12 months of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in

' accordance with Section 13 of ANSI N510-1980 meets the laboratory testing criterion of greater than or equal to 90%

efficiency when tested with methyl iodide at 80*C and 70%

relative humidity.

c. At least once per 18 months by verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the main purge system.

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FA Y-UNIT 2 3/4 9-17 AMEN 0 MENT NO.37

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' REFUELfNG OPERAT80NS i

SURVEILLANCE REQUIREMENTS (Continued)  !

d. After each complete or partial replacement of a HEPA filter bank by verifying-that the HEPA filter banks remove greater l than or equal to 99.5% of the DOP when they are tested in-place in accordance with Section 10 of ANSI N510-1980.

e. After each complete or partical replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.5% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with Section 12 of ANSI M510-1980 while operating the main purge system.

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3/4 9-18 AMENDMENT NO.37 hgLEY-UNIT 2

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. -. 3 -- :::an SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0. 4s TO FACILITY OPERATING LICENSE N0. NPF-2 AND AMENDMENT NO. 37 TO FACILITY OPERATING LICENSE NO. NPF-8 ALABAMA POWER COMPANY JOSEPH M. FARLEY NUCLEAR PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-348 AND 50-364 Introduction By letter dated March 4,1983 Alabama Power Company (APCo) proposed changes to the Technical Specifications relating to charcoal filters. The NRC staff .

did not consider that APCo had sufficiently justified the requested changes.

During the course of several telecon discussions in June, July and August 1983, we advised APCo of our concerns relating to the testing criteria being j proposed.  ;

Subsequently, by letter dated March 1,1984, in response to the NRC staff concerns, APCo provided modifications.to the original proposal along with a k more detailed bases for the purposed changes. Our discussion and evaluation follows.

j - Discussion Certain banks of charcoal filters are used to absorb the airborne radioactivity following a postulated loss-of-coolant-accident (LOCA). The Control Room Emergency Air Filtration System and the Penetration Room Air l Filtration System both contain charcoal filter banks to assure that the radiation exposures to personnel would remain within guidelines of 10 CFR 50, Appendix A, General Design Criteria 19. Also, the containment purge exhaust filter assures that any airborne radioactivety resulting from a postulated fuel handling accident during refueling would be absorbed prior to reaching the environment. Technical Specification surveillance requirements are necessary to assure that licensees use Comission approved testing methods and criteria for testing the charcoal filter radioactivity absorber's efficiency.

Evaluation The originally issued Technical Specifications for the Farley Nuclear Plant referenced the analysis techniques and acceptance criteria of Regulatory Guide (RG) 1.52. Revision 2. March 1978. These references may have led to misinterpretations of test methods and efficiency requirements as evidenced in Licensee Event Report 83-006, an event which occurred on February 15, i 1983. For these reasons APCo proposed changes to the Technical

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Specifications by letter dated March 4, 1983, supplemented March 1, 1984, which we have evaluated. Briefly stated the changes would:

(1) lower the HEPA and char

• coal . filter system surveillance leak I test acceptance requirement from 99.951 (RG 1.97) to 99.5% removal I efficiency, (2) specify specific laboratory charcoal testing methyl iodide removal efficiencies that are consistent for iodine removal credit allowed by the staff and specified in the Final Safety Evaluation Report, and (3) specify the latest NRC staff approved testing methods to be used for performing HEPA and charcoal filter leak testing and also charcoal filter laboratory methyl iodide testing.

Our review indicates that the overall iodine removal efficiency, as shown in the enclosed Table is above the iodine removal credit considered in the NRC

. staff Safety Evaluation when the license was granted. Therefore, new Technical Specifications as proposed in the March 1,1984, APCo letter are acceptable on this basis.

Safety Sumary On the basis of o'ur review we conclude that these Technical Specification changes would result in no significant fncrease in accident-related site boundary doses from doses detennined in'the earlier analysis reported in the Farley, Unit Nos. I and 2, Safety Evaluation when the plants were licensed.

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Environmental Consideration This amendment involves a change in the installation or use of a facility component located within the restricted area. The staff has determined that the amendment involves no significant increase in the amounts of any effluents that may be released offsite and that there is no significant i

increase in individual or cumulative occupation radiation exposure. The Comission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public coment on such finding. Accordingly, this amendment meets the eli criteria for categorical exclusion set forth in 10 CFR Sec 51.22(c)gibility (9).

Pursuant to 10 CFR 51.22(b) no entironmental impact statement or environmental asses,sment need be prepared in connection with the issuance of this amencment.

Conclusion ,

We have concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Comission's regulations and the issuance of these amendments will not be inimical to the comon defense and security or to the health and safety of the public.

( Dated: June 22, 1984 Principal Contributors:

. R.\ Fell E.%eeves

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• ATTACHMENT TABLE J

' o ~ OVERALL ESF FILTER SYSTEM

' IODINE REMOVAL EFFICIENCY Iodine Removal Credit Allowed by Staff Tech Spec Tech Spec Overall for Organic Methyl Iodide Iodine removal and Elemental Leak Test Efficiency

• Iodine

! Filter System Efficiency Test Efficiency Control Room

> 99.32% 9 95 Inlet (with > 99.5%

> 99.825%

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heaters)

),98.50% 95%

Recirculation > 99.5%

> 99.0%

> 99.5% > 95.0% > 94.42% 90% for Penet ration ~ ~ -

Elemental

.BggyL(Fue1 and 70% for Handling Organic Accident and LOCA) (no

heaters) -

> 90.0% > 89.55% 90% for i Containment > 99.5% "~

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- Elemental and

(. Purae Exhaust 70% for Organic (Fuel Handling (Unit 1)

, Accident 30% Organic

. Inside (Unit 2)

Containment)

(no heaters) l

• Calculated removal efficiency for Organic Iodine; Elemental iodine should be greater.

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Enclosure 11 3/4.8 ELECTRICAL POWER SYSTEMS 4'7 4

3/4.8.1 A.C. SOURCES A.C. SOURCES - OP RATING LIMITING CONDITION FOR OPERATION 3.8.1.1 Asamibimum,thefollowingA.C.electricalpowersourcesshallbe OPERABLE:

a. Two physically independent circuits between the offsite transmission network and the onsite Class IE distribution system, and

b. Three separate and independent diesel generators, each with:

1. A separate day fuel tank containing a minimum of 225 gallons of fuel for Div 1 and Div 2 and 204 gallons of fuel for Div 3,

2. A separate fuel storage system containing a minimum of 69,430 gallons of fuel for Div 1 and Div 2 and 34,824 gallons of fuel for Div 3, and 4

3. A separate fuel transfer pump.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a. With one offsite circuit of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C.

sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter. If either diesel generator Div 1 or Div 2 has not been successfully tested within the past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, demonstrate its OPERABILITY by performing Surveillance Re-quirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 for each such diesel-gen-erator separately within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Restore the offsite circuit to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. With either diesel generator Div 1 or Div 2 inoperable, demonstrate the OPERABILITY of the above required _A.C. offsite sources by perform-ing Surveillance Requiremert 4.8.1.1.1.a within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter. If the diesel generator became inoperable due to any cause other than preplanned preventive maintenance or testing, demonstrate the OPERABILITY of the remaining OPERABLE diesel generators by performing Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 separately for each diesel generator within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />";

• This test is required to be completed regardless of when the inoperable diesel generator is restored to OPERABILITY. The provisions of Specification 3.0.2 re not applicable.

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E 5a v ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

-ACTION (Continued)

f. With both of the above required offsite circuits inoperable, demon-strate the OPERABILITY of three diesel generators by performing Sur-viellance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 separately for .

each diesel generator within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> unless the diesel generators are already operating; restore at least one of the above required offsite circuits 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 be in at least HOT SHUTDOWN within the ne73; 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. With only one offsite circuit restored to OPERABLE status, restore at least two offsite circuits to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time of initial loss or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN 4

within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. A successful test (s) of diesel gener-ator OPERABILITY per Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 performed under this ACTION statement for the OPERABLE diesel generators satisfies the diesel generator test requirements of ACTION a.

g. With diesel generators Div 1 and Div 2 of the above required A.C.

electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirement l 4.8.1.1.1.a within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter and Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 for diesel generator Div 3 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> *. Restore at least one of the inoperable diesel generators Div 1 and Div 2 to OPERABLE status within

. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Restore both diesel generators Div 1 and Div 2 to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time of initial loss or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

h. With one offsite circuit of the above required A.C. electrical power sources and diesel generator Div 3 inoperable, apply the requirements of ACTION a and d specified above.

i. With either diesel generator Div 1 or Div 2 inoperable and diesel gen-erator Div 3 inoperable, apply the requirements of ACTION b, d and e specified above.

• This test is required to be completed regardless of when the inoperable diesel generator is restored to OPERABILITY. The provisions of Specification 3.0.2 are not applicable.

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>m ELECTRICAL POWER SYSTEMS O

SURVEILLANCE REQUIREMENTS (Continued)

7. Verifying the pressure in all diesel generator air start receivers to be greater than or equal to 210 psig.

b. At least once per 31 days and after each operation of the diesel where the period of operation was greater than or equal to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> by checking for and removing accumulated water from the day tank.

c. At least once per 92 days by checking for and removing accumulated water from the fuel oil storage tanks.

d. At least once per 92 days and from new fuel oil prior to its addi-tion to the storage tanks by verifying that a sample obtained in accordance with ASTM-0270-1975 meets the following minimum require-ments in accordance with the tests specified in ASTM-D975-1977:

1) A water and sediment content of less than or equal to 0.05 volume percent;

2) A saybolt universal viscosity at 100*F of greater than or equal to 32.6 sus, but less than or equal to 40.1 sus;

3) An API gravity as specified by the manufacturer at 60*F of greater than or equal to 26 degrees, but less than or equal to 36 degrees;

! 4) An impurity level of less than 2 mg of insolubles per 100 ml

,s- when tested in accordance with ASTM-02274-70; analysis shall be completed within 7 days after obtaining the sample but may be sampled and analyzed after the addition of new fuel oil; and

5) The other properties specified in Table 1 of ASTM-0975-1977 ahd Regulatory Guide 1.137, Revision 1, October 1979, Position

, 2.a., when tested in accordance with ASTM-0975-1977; analysis l

shall be completed within 14 days after obtaining the sample l

but may be sampled and analyzed after the addition of new fuel

' oil,

e. At least once per 18 months *, during shutdown, by:

1. Subjecting the diesel to an inspection in accordance with instructions prepared in conjunction with its manufacturer's recommendations for this class of standby service.

2. Verifying the diesel generator capability to reject a load of greater than or equal to 1400 kw (LPCS pump) for diesel generator Div 1, greater than or equal to 725 kw (RHR 8 pump or RHR C pump) l I

~

• For any start of a diesel, the diesel must be loaded in accordance with the g manufacturer's recommendations.

\

PbFY-UNIT 1 3/4 8-5

\

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).

ELECTRICAL POWER SYSTEMS a

SURVEILLANCE REQUIREMENTS (Continued) .

state voltage and frequency of the emergency bus shall be maintained at 4160 420 volts and 60 1 1.2 Hz during this test.

5. Verifying that on an ECCS actuation test signal, without loss of offsite power, the diesel generator starts

• on the auto-start signal and operates on standby for greater than or equal to 5 minutes. The generator voltage and frequency shall be 4160 420 volts and 60 1.2 Hz within 10 seconds after the auto-start signal for Div 1 and Div 2 and within 13 seconds after the auto-start. signal for Div 3; the steady state generator voltage

, and frequency shall be maintained within these limits during this test.

6. Simulating a loss of offsite power in conjunction with an ECCS actuation test signal, and:

a) For divisions 1 and 2:

1) Verifying de-energization of the emerger!cy busses and load shedding from the emergency busses.

2) Verifying the diesel generator starts

• on the auto-start signal, energizes the emargency busses with permanently connected loads within 10 seconds, energizes the auto-connected emergency loads and operates for greater than or equal to 5 minutes while its generator is

. loaded with the emergency loads. After energization, the steady state voltage and frequency of the emergency busses shall be maintained at 4160 t 420 volts and 60 2 1.2 Hz during this test.

b) For division 3:

1) Verifying de-energization of the emergency bus.

2) Verifying the diesel generator starts

• on the auto-start signal, energizes the emergency bus with its loads and the auto-connected emergency loads within 13 seconds and operates for greater than or equal to 5 minutes while its generator is loaded with the emer-gency loads. After energization, the steady state i

• All diesel generator starts for the purpose of this Surveillance Requirement may be preceded by an engine prelube period. The diesel generator start (10 sec)/ load (60 sec) from ambient conditions shall be performed at least once per 184 days in these surveillance tests. All other engine starts for the purpose of this surveillance testing may be preceded by other warmup pro-cedures recommended by the manufacturer so that the mechanical stress and wear on the diesel engine is minimized.

g -

PdLRY-UNIT 1 3/4 8-7

\

)4 j ELECTRICAL POWER SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) to standby. operation, and (2) automatically energizes the emer-gency loads with offsite power.

12. Verifying that each fuel transfer pump transfers fuel from the fuel storage tank'to the day tank of each diesel.

13. Verifying that the automatic load sequence timers are OPERABLE with the interval between each load block within i 10% of its design interval for diesel generators Div 1 and Div 2.

14. Verifying that the following diesel generator lockout features prevent diesel generator starting only when required: ,

a. For diesel generators Div 1.and Div 2:

1) Control room switch in pull-to-lock (with local / remote switch in remote).

2) Local / remote switch in local

3) Barring device engaged

4) Inop/ Normal switch in inop

b. For diesel generator Div 3:

1) Emergency run/stop switch in stop

2) Maintenance / auto / test switch in maintenance-

f. At least once per 10 years or after any modifications which could'af-fect diesel generator interdependence by starting all three diesel generators simultaneously, during shutdown, and verifying that all three diesel generators accelerate to at least 441 rpm for diesel generators Div 1 and Div 2 and 882 rpm for diesel generator Div 3 in less than or equal to 10 seconds.

g. At least once per 10 years by:

1. Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite or equivalent solution, and

2. Performing a pressure test of those portions of the diesel fuel i oil system designed to Section III, subsection ND of the ASME Code in accordance with ASME Code Section 11 Article IWD-5000.

4.8.1.1.3 Reports - All diesel generator failures, valid or non-valid, shall be reported to the Commission pursuant to Specification 6.9.2 within 30 days.

Reports of diesel generator failures shall include the information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1 August 1977.

j If the number of failures in the last 100 valid tests, on a per nuclear unit i basis, is greater than or equal to 7, the report shall be supplemented to in-telude the additional information recommended in Regulatory Position C.3.b of gulatory Guide 1.108, Revision 1, August 1977.

PEhY-UNIT 1 3/4 8-9

- - - - , . . _ _- ___m

e 5

ke- dic (har l Enclosure 12 ts?

l REFUELING OPERATION $

3/4.9.6 REFUELING MACHINE LIMITING CONDITION FOR OPERATION 3.9.6 The refueling eachine and auxiliary hoist shall be used for movement of drive rods or fuel assemblies and shall be OPERABLE with:

a. The refueling machine used for movement of fuel assemblies having:

1) A minious capacity of 4000 pounds, and

2) An overload cutoff liett less than or equal to 3900 pounds.

b. The auxiliary hoist used for latching and unlatching drive rods having:

1) A minious capacity of 3000 pounds, and ,

2) A load indicator which shall be used revent lifting loads in excess of 1000 pounds.

APPLICABILITYi During movement of drive rods or fuel asseublies within the reactor vessel.

M: .

With the requirements for/falib and/or hoist OPIRABILITY not satisfied, suspend use of any inoperable G==Maui== ' - and/or auxiliary hoist from operations involving vessel. the movement of${# rive' rods endJuei assentlieswithin th

$URVEILLANCE REQUIREMENTS g

NM 4.9.6.1 leen nu used for movement of fuel assemblies within r

the reactor vessel shall be demonstrated OPERABLE within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of such operations by performing a load test of at least 4000 pounds and demonstrating an automatic load cutoff h the crunt load exceed /ing 3900 pounds, nb 4.9.6.2 auxiliary hoist and associated load indicator used for movement of drive rods within the reactor vessel shall be demonstrated OPERABLE within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of such operations by performing a load test of at least 3000' pounds.

l2,SD MILLSTONE - UNIT 3 3/4 9-6

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I COMPARISON OF THE TECHNICAL SPECIFICATIONS WITH THE S R l,/,

SECTION PACE SUBJECT REMARKS 3.4.1 3-7 Water tight doors which serve fire Water tight doors which also provide fire and flood protection protection will be maintained as part of the Fire l Protection Program. CPC has proposed to eliminate Technical Specifications pertaining to fire detection and suppression systems and fire barriers. Our proposal would make the FSAR the controlling document for the Fire Protection Program with certain administrative controls implemented via Section 6.0 of the Technical Specifications.

j 3.5.1.3 3-19 Inspection and testing of turbine valves 'Ihis is addressed in Specification 3/4.3.4.

i 3.7.4 3-30 Seismic instrumentation inservice The seismic instrumentation inservice surveil-surveillance program lance program is described in 4.3.3.3 of the Technical Specifications.

3.8.1 3-32 Surveillance requirements for contain- This is addressed in Specification 3/4.6.1.6.

ment structural integrity 3.9.6 3-49 Limiting conditions for operation and Limiting conditions for operation and surveil-leak rate testing program for RCS lance requirements for RCS pressure isolation 9 ,

pressure isolation valves valves are provided in 3/4.4.6.2 of the Technical i

i Specifications.

! 4.3.2 4-22 Peaking Factor Limit Axial flux difference is limited to + 5% for less than or equal to 3000 MWD /MTU burnup as per Specification 3/4.2.1.

4.3.2 4-24 Regulating rod insertion Control rod insertion limits are specified in 3/4.1.3.6 of the Tec.hnical Specifications. g 2.

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j ./ COMPARISON OF THE TECIGIICAL SPECIFICATIONS WITH THE SER (Cont'd)

SUBJECT REMARKS i SECTION PACE 4-30 Minimum required reactor coolant flow Minimum required RCS flow is specified in Table

, 4.4.3.2 i 3.2-1, Specification 3.2.5.

Thermal margin used to offset the rod This is addressed in the bases for Specifications 4.4.4.1 4-31

' bow penalty 3/4.2.2 and 3/4.2.3.4.4.6 4.4.6 4-32 N-1 loop operation during modes 1 and 2 Specification 3/4.4.1 requires that all reactor j

is to be prohibited. (N-1 loop operation coolant loops be in operation during Modes 1 and 2.

j All references to N-1 operation which appear in

< is operating with one cooling loop out of se rvice.) the standard Technical Specifications have been i

deleted from the VEGP Draft Tech. Specs.

i 5-5 PORV setpoint values to be provided'by The PORV setpoint values have been provided as 5.2.2.2 Figure 3.4-4 of Specification 3/4.4.9.3.and will-5-6 ' applicant and periodically upd,ated.

i '

be periodically updated as discussed in the basis l

for 3/4.4.9 under Cold Overpressure Protection Systems.

50 degree fahrenheit mismatch between This is addressed in Specification 3/4.4.1.3 and 5.2.2.2 3/4.4.1.4.

i RCS (2500F) and the secondary side I of the steam generators (3000F).

5.2.2.2 Technical Specifications on pressure The low temperature overpressurization concerns and the maximum temperature mismatch have been incorporated into Specifications 5.2.2.3 3/4.1.2.2, 4.1.2.3.2, 3.1.2.4, 4.1.2.4.2, between the secondary and the primary system before a reactor coolant pump 3.4.9.3, 3.8.1.2, 3.8.2.2, 3.8.3.2, 3.5.2, I

l may be started. Incorporation of cold 3/4.5.3.1, and 3/4.5.3.2.

j overpressurization analysis results

' into the Technical Specifications.

5.2.3 5-8 Chemical control of the primary coolant , This concern is addressed in Specification 3/4.4.7.

j water.

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Page 3 of 11

,.' COMPARISON OF THE TECIDf1 CAL SPECIFICATIONS WITH -THE SEE (Cont'd)

SUBJECT REMARKS SECTION PAGE 5-23 The staff considers the examination of inservice inspection of steam generator tubes is -

5.4.2.2.2 addressed in Specification 3/4.4.5.

the steam generators a confirmatory issue, contingent upon the applicant conforming with the requirements of the applicable provisions of the Standard Technical Specifications.

The limitation of RCS cooldown to 100 This is addressed in Specification 3/4.4.9.1.

5.4.7 5-25 degrees fahrenheit per hour due to thermal stress considerations.

Applicant must adopt operability require- This is addressed in Specification 3/4.4.11.

5.4.12 5-34 ments before the vent system is considered fully acceptable. (license condition) .

The Technical Specifications must include This is addressed in Specification 3/4.6.2.2.

6.1.1 6-3 surveillance requirements to verify that the 30 weight percent sodium hydroxide is maintained.

The applicant assumed an initial contain- Our February 28, 1986, submittal shows a pressure 6.2.1.1.1 6-6 of A5 psig- for Pa which addresses this concern.

ment pressure of 15 PSIA which is less than the Technical Specification limit of 17.7 PSIA during normal operation. The applicant has proposed adding the dif ference (2.7 psig) to the peak. calculated pressure (41.9 psig) to determine the test pressure (Pa) for containment leakage testing.

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/ COMPARISON OF THE TECHNICAL SPECIFICATIONS WITH THE SER (Cont'd)

PAGE SUBJECT REMARKS i SECTION l' 6.2.4 6-13 The 24-inch preaccess purge lines are This is addressed in Specification 3/4.6.1.7.

sealed closed during operational Modes 1, 2, 3, and 4 and are verified to be closed at least every 31 days.

This requirement is to be included in the plant's Technical Specifications.

6-16 Air locks opened during periods when This is addressed in Specification 4.6.1.3.

6.2.7 containment integrity is not required by the plant's Technical Specifications

shall be tested at the end of such a -

! period at not less than Pa.

6-17 Exemption from the requirement of This is addressed in Specification 4.6.1.3.

)

4 6.2.6 10 CFR 50 Appendix J, paragraph III.D.2 is justified and acceptable for Vogtle Units 1 and 2 and appropriate require-I ments will be added to the plant Technical Specifications.

6.2.6 6-17 The plant's Technical Specifications will This is addressed in Specifications 3/4.6.1.1, i

contain appropriate surveillance require- 3/4.6.1.2, 3/4.6.1.3 and 3/4.6.1.7.

I ments for containment leak testing includ-ing test frequencies.

1 6.3.6 6-31 Testing of the operability of the ECCS. This is addressed in Section 3/4.5 of the Technical Specifications.

i 7.2.2.1 7-7 Time constants for signal compensations This is addressed in Specification 3/4.3.2,

' are adjustable setpoints within the ana- Table 3.3-4 and Specification 2.2.1, Table 2.2-1.

log portion of the safety system and are

• to be incorporated into the Technical Specifications.

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Page 5 of 11

,/

./ COMPARISON OF THE TECHNICAL SPECIFICATIONS WITH THE SEE (Cont'd)

PACE SUBJECT REMARKS SECTION 7.2.2.2 7-8 The staff will include in the plant Tech- Turbine trip on reactor trip is initiated by.

nical Specifications a requirement to interlock P-4. Operability and surveillance re-periodically test circuits involved in quirements are addressed in Tables 3.3-3 and 4.3-2 the turbine trip following a reactor trip. of Specification 3/4.3.2.

7.2.2.4 7-9 In performing its reviews the staff The Westinghouse Statistical Setpoint Methodology audits detailed information on the will be utilized and a plant-specific report will methodology used to establish the Tech- be provided for NRC staff review.

nical Specification trip setpoints and allowable values for the reactor pro-tection system assumed to operate in the FSAR accident and transient analysis.

7.2.2.5 7-9 The specific power level setpoint below We setpoint for the P-9 interlock is specified which a reactor trip following a turbine in Specification 2.2.1, Table 2.2-1.

trip is blocked will be reviewed and specified in the plant Technical Specifi-cations..

7.3.3.1 7-20 he initiation signals and circuits of the Testing for APWS initiation signals and circuits AFWS are testable during power operation are specified in Specification 4.3.2.1, and the test requirements will be in- Table 4.3-2; and Specification 4.3.2.2.

cluded in the plant Technical Specifica-ions.

7.3.3.3 7-21 Until an acceptable circuit modification The modification has been presented to the NRC is installed the staff will require plant and has been implemented in the design (response Technical Specifications to include month- to Open Item 59). This item remains in the SER ly (in lieu of quarterly) testing of slave until the NRC has determined by inspection that relays.

the modification has been installed.

4

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_ _ . _ _ _ _ _ . . _ _ _ . _ . _ _ _ _ . _ _ . . _ _ . _ . . _ _ _ _ _ _ _ _ . ~ . . . _ _ _ . _ _. _ _ _ _ . .,

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l *,. COMPARISON OF THE TECHNICAL SPECIFICATIONS WITH THE SER (Cont'd)

.i i

PAGE -SUBJECT REMARKS SECTION 1

7.3.3.4 7-21 The staff will ensure that any environment- This will be addressed in the Statistical Setpoint al errors resulting from environmental Study.

l temperature effects on level instrument reference legs are taken into account l

i during review of setpoint methodology

! and plant Technical Specifications.

I 7.6.2.3 7-37 Periodic surveillance of safety functions This is addressed in Specifications 3/4.5.4, l in FSAR Sections 7.6.6.1 through 7.6.6.4 3/4.7.12 and 3/4.3.3.11.

4

! and 7.6.6.6 and their instrument channels.

1 I 8.3.1 8-4 Because the manual transfer scheme could This item remains open pending demonstration that I result in the connection of both Class 1E one reserve auxiliary transformer has the necessary trains to one reserve auxiliary transfore- capacity.

l l er a Technical Specification limitation -!

l will be needed in order to preclude this i

} from happening unless the applicant can show that one reserve auxiliary transformer l i has sufficient capacity to start and run the loads of both Class 1E trains.

8.4.3 8-11 The applicant has stated that the Vogtle In our February 28, 1986, submittal, we have Technical Specifications will explicitly proposed that Table 3.8-2 of Specification identify the MOVs which are required to 3/4.8.4.2.be deleted and that'the list of valves-have thermal overload bypass jumpers be controlled via plant procedures.

in place during plant operation.

8.4.4 8-11 BTP ICSB 18 (PSB) requires for valves These valves are addressed'in Surveillance that require power lockout to meet the Requirement 4.5.2.a.

single failure criterion.in the fluid systems that the valves and their re-

• quired position be listed in the Tech-nical Specifications.

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/ CnMPARISON OF THE TECHNICAL SPECIFICATIONS WITH THE SEE (Cont'd)

SECTION PACE SUBJECT REMARKS 8.4.5 8-14 The staff will confirm that the overcurrent These breakers are addressed in Specification settings of the circuit breakers protecting 3/4.8.4.1.

class 1E circuits from non-Class lE loads are tested periodically.

9.2.2.1 9-15 Availability of CCW pumps not running will Surveillance requirements and operability of the be assured by periodic test and inspections CCW system are addressed in Specification 3/4.7.3.

as required by the plant Technical Speci-fications.

9.2.5 9-21 Components of the UHS which are not norm- Surveillance requirements and operability of the ally operating will be tested in accord- UHS are addressed in Specification 3/4.7.5.

ance with plant Technical Specirfications.

9.2.6 9-23 The condensate storage facility is normally Surveillance requirements and operability of the in operation and its safety function condensate storage facility are addressed in (supply to AFW system) is functionally Specification 3/4.7.1.3.

tested in accordance with the plant Technical Specifications.

9.4.1 9-37 With regard to Unit 1 operation while As shown in FSAR Figure 9.4.1-2 (Sheet 2 of 3) a Unit 2 is still under construction, the portion of the common HVAC ducting in the control staff will review the Technical Specifi- room will be physically removed and the ends capped cations for Unit 1 to ensure complete to ensure complete isolation of the Unit 1 and isolation from the Unit 2 HVAC system Unit 2.HVAC systems. Based on the complete before an operating license is issued separation provided by not installing a portion of for Unit 1. the ducts, we feel that there is no need for a technical specification to ensure isolation.

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Page 8 of 11 g'-

COMPARISON OF THE TECHNICAL SPECIFICATIONS WITH THE SEE (Cont'd)

L SECTION PAGE SUBJECT REMARKS

[

J i 9.5.4.2 9-65 In FSAR Amendment 16 the applicant This is addressed in Surveillance Requirements i stated that the Vogtle fuel oil sur- 4.8.1.1.2.c, 4.8.1.1.2.d, and 4.8.1.1.2e.

veillance program would follow the program that was approved for the McGuire Nuclear Station. The appli-cant further stated that this program would be specified in the Vogtle Tech Spec.

I 10.3.5 10-7 Because the corrosion phenomena involved This is addressed in Specification 6.8.4.c.

in steam generator degradation are so j compler and, considering the state of the art as it exists today, the staff j finds that, instead of specifying limiting conditions in the plant Technical Specifications, a more effective approach would be to specify a Technical Specification that required the implementation of a secondary water chemistry monitoring and control program containing appropriate procedures and administrative controls.

10.4.7 10-16 The safety related portions of the The safety-related portions of the Condensate and Condensate and Feedwater system are Feedwater System are inspected as required by located in accessible areas and will Specification 4.0.5, operability and surveillance receive periodic inspection and test- of the main feedwater isolation function is ing in accordance with the Technical addressed in Specification 3/4.3.2, and opera-Specifications, bility and surveillance of the AFWS is addressed l 'in Specification 3/4.7.1.2 as well as 4.0.5.

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j ./ COMPARISON OF THE TECIDfICAL SPECIFICATIONS WITH THE SEE (Cont'd)

I SUBJECT REMARKS SECTION PAGE 1

10-20 The applicant will perform monthly Operability, maintenance outage times, and 3 10.4.9 surveillance of the AFWS is addressed in tests in conformance with the Standard Technical Specifications for Westinghouse Specification 3/4.7.1.2.

1 pressurized water reactors NUREG-0452.

1 l

The applicant has indicated that the

outage time limit and the subsequent action time in the Technical Specifica-I tions will be as required by the I Standard Technical Specifications.

I The zoning system and the access control Section 12.3.1.2 of the VEGP FSAR states that, "The 12.3.1 12-4 "

features will also meet the posting entry posting of radiation signs, control of personnel requirements of 10 CFR 20.203 or access, and use of alarms and locks'are in com-l Standard Technical Specifications and pliance with requirements of 10 CFR 20.203." This are consistent with R.G. 8.8. is addressed in Section 6.12 of the Technical Specifications.

The fire team manning requirements will As has been stated previously, GPC has proposed l 13.1.2 13-8 that the fire protection elements of the Technical j

be met without impacting the minimum ,

onshift operating staff requirements as Specifications be deleted and that the fire protec-  !

j described in the Vogtle Technical Speci- tion program be controlled administrative 1y via the ,

the FSAR.

! fications.

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,/ CONPARISON OF THE TECHNICAL SPECIFICATIONS WITN THE SER (Cont'd)

PAGE SUBJECT REMARKS SECTION 13.4.1

~

The PRB will review certa 1n procedures, The responsibilities of the PRB are addressed in tests and experiment changes to procedures, Specification 6.5.1.6.

equipment and systems that involve an unreviewed safety question, proposed changes to the Technical Specifications, reportable events, the Security Plan, the Emergency Plan, and other items related to the operation of the plant.

Additionally, the subject of proposed changes or modifications to systems or equipment and investigations of violations of the Technical Specifications have not been adequately addressed. (open ites.

13-24 The SRB will consist of a Chairman and a This is addressed in Specification 6.5.2.

13.4.2 minimum of six members and will function to provide an independent offsite review of activities as described in Section 6.5.2 of the Standard Technical Specifi-cations.

15.1.5.2 15.10 The staff will review the Vogtle plant's The limits on iodine concentration in the reactor 15.6.2 Technical Specifications to ensure that coolant are addressed in Specification 3/4.4.8 the limits for iodine concentration in and limits on primary-to-secondary leakage are the reactor coolants (consisting of a addressed in Specification 3/4.4.6.2.

maximum allowable limit and an equil-ibrium limit) are incorporated.

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,/ COMPARISON OF THE TECIBIICAL SPECIFICATIONS WITH THE SEE (Cost'd) 1 SUBJECT REMARKS SECTION PAGE 15-19 In FSAR Amendment 13 the applicant com- This is addressed in Specification 4.9.1.3 15.4.6 mitted to include in its Technical Speci-j fications a requirement to lock closed l

valves 175, 176, 177, and 183 during refueling. The staff will confirm that this requirement is included in the Technical Specifications.

i

! 15.4.8.2 15-22 A maximian of 1-gpa primary-to-secondary This is addressed in Specification 3/4.4.6.2 i 15.1.5.2 leak rate is assumed (as limited by Technical Specifications).

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