L-99-030, Forwards SNC Review Comments on Draft SE & marked-up Copy of Draft SE Incorporating SNC Comments Re Proposed Conversion to ITS
| ML20211K380 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 08/30/1999 |
| From: | Dennis Morey SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NEL-99-0307, NUDOCS 9909070206 | |
| Download: ML20211K380 (49) | |
Text
4' 4
7.
Dave Morey Sxth:rn Nuclear
. Vice President -
Op: rating C:mpany,Inc.
Farley Project Post Office Box 1295 Birmingham Alabama 35201 Tel 205 992.5131 4
SOUTHERN August 30, 1999-COMPANY Energy to Serve )*ourWorld*
Docket Nos.:
50-348 NEL-99-0307 50-364
- U. S. Nuclear Regulatory Commission A'ITN: Document Control Desk Washington, DC 20555-0001 Joseph M. Farley Nuclear Plant Draft Safety Evaluation for Proposed Conversion to the Imoroved Technical Soccifications - SNC Review Comments Ladies and Gentlemen:
NRC letter dated August 6,1999, provided a draft safety evaluation (SE) for the proposed conversion to the Improved Technical Specifications for Farley Nuclear Plant (FNP), Units 1 and 2.
The purpose of this letter is to provide SNC comments on that draft SE.
Attachment I provides the SNC review comments on the draft SE. Attachment II provides a marked-up copy of the draft SE incorporating the SNC comments. Two sections of the draft SE
- were not included in the August 6,1999 transmittal. The draft SE states that the SE related to the change in RCP flywheel inspection frequency will be provided at a later date. In addition, the Current Technical Specifications (CTS) Discussion of Change Tables were not included with the attached draft SE. The Tables were received by SNC on Friday, August 27,1999. SNC will review the Tables and provide comments.
If there are any questions, please advise.
i Respectfully submitted, 9% '7?w
- /
Dave Morey
/
J g0038 g) c'
/
.L
/
WAS/maf:SER_ review. DOC Attachments:
I.
SNC Review Comments - Joseph M. Farley Nuclear Plant, Units I and 2, Draft Safety Evaluation for Proposed Conversion to the Improved Technical Specifications II.
SNC Markup - Joseph M. Farley Nuclear Plant, Units I and 2, Draft Safety Evaluation for Proposed Conversion to the Improved Technical Specifications 9909070206 990830 PDR A00CK 05000348 F
PDR L
q 1
q 4
Page 2 U. S. Nuclear Regulatory Commission cc:
Southem Nuclear Operatine Company Mr. L. M. Stinson, General Manager -
i U. S. Nuclear Reculatory Commission. Washincton. D. C.
Mr. L. M. Fadovan, Licensing Project Manager - Farley
)
U. S. Nuclear Regulatory Commission. Recion II Mr. L. A. Reyes, Regional Administrator Mr. T. P. Johnson, Senior Resident Inspector - Farley I
o ATTACIIMENT I SNC REVIEW COMMENTS JOSEPII M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2, DRAFT SAFETY EVALUATION FOR PROPOSED CONVERSION TO TIIE IMPROVED TECIINICAL SPECIFICATIONS l
1
i i
DRAFT SAFETY EVALUATION FOR PROPOSED CONVERSION TO IMPROVED STANDARD TECHNICAL SPECIFICATIONS SNC REVIEW COMMENTS i
Page Paragraph Change 1
1 Date for Unit 1 original operating license should be June 25,1977.
I 1
An additionalletter was transmitted on August 19,1999. This letter should be referenced.
1 2
Capitalize first words in names for TS (Improved, Current, Standard).
1 3
Meetings and conference calls have continued past April 20,1999.
Need to reflect this. The final date should be inserted.
I 5
8 Item 6 is renumbered as item 7 and a new item 6 is insened for generic less restrictive changes (LC).
7 1
The phrase " Categories I through VII" should be changed to
" Categories I through VIII."
l 8
1 The phrase in the first line of the first paragraph on the page should be changed from " unavailability due to test" to " unavailability due to testing."
8 2
The phrase in the first line of the second paragraph on the page should be changed from " unavailability due to test" to
" unavailability due to testing."
10 1
On the 18th line of the first paragraph on the page (under the heading " Deletion of SR (Category VII)"), the word "be" should be I
deleted.
l 10 1
The following sentence should be added to the end of the first paragraph under the heading " Deletion of SR (Category VII)," to be consistent with the other category discussions:"These changes are consistent with the STS and changes specified as Categon VII l
are acceptable."
10 3
In the second paragraph under the heading " Deletion of Requirement for 30-day Special Report to NRC," the phrase "Categoy I through IX" should be changed to " Category I through VIII."
11 8
"UFSAR" should be changed to "FSAR" for consistency (four places).
j 12 1,2, and 3 "UFS AR" should be changed to "FS AR" for consistency (five places).
l 12 4
"USAR" should be changed to "FS AR" for consistency (two places).
12 4
The first sentence of the first paragraph under the heading
" Administrative Requirements Redundant to Regulations," should include deletion of regulations that are redundant to CFR or Regulatog Guide requirements (requirement is already in the CFR or Regulatory Guide - see RAI 3.1.3-5 response).
13 1
Revise the first sentence of the paragraph to state:". relocated to
)
Page1of4
r l
DRAFT SAFETY EVALUATION FOR PROPOSED CONVERSION l
TO IMPROVED STANDARD TECHNICAL SPECIFICATIONS l
SNC REVIEW COMMENTS Page Paragraph Change l
SNC-controlled documents during the ITS conversion or deleted."
13 3
The second sentence of the paragraph should include deletion of regulations that are redundant to CFR or Regulatory Guide requirements (requirement is already in the CFR or Regulatory l
Guide - see RAI 3.1.3-5 response).
I 13 5
"UFS AR" should be changed to "FS AR" for consistency.
13 7
"UFSAR" should be changed to "FS AR" for consistency.
13 7
"QA Plans" should be changed to "QA Program." Also, additional documents should be added to the list (see March 12,1998 letter, ).
13 10 "UFS AR" should be changed to "FS AR" for consistency.
- 15. _
3..
The phrase in the last sentence of the first paragraph under the heading " CTS 3/4.3.3.2 -Movable Incore Detectors," should be l
revised from "However, the movable incore detectors the do not..."
to "However, the movable incore detectors do not..."
16 1
TS 3/4.3.4 should be changed to CTS 3/4.3.4 for consistency.
16 2
The FS AR reference for the turbine overspeed protection testing program is correct (FSAR section 10.2.2). The brackets and question marks should be deleted.
16 4
TS 3/4.4.8 should be changed to CTS 3/4.4.8 for consistency.
17 2
The paragraph should be revised as follows for clarity: "....Since the pressurizer normally operates in temperature ranges above those for which there is a reason for concern of nonductile failure, temperature limits are placed on the pressurizer to assure compatibility of operation with the fatigue analysis performed in i
accordance with the ASME Code requirements. However, a failure i
of pressurizer integrity would result in an analyzed event (loss of coolant accident) for which numerous systems and components are required and retained in the TS. While these limits represent l
operating restrictions and Criterion 2 includes operating restrictions, Criterion 2 applies only to those operating restrictions required to preclude unanalyzed accidents and transients.
Therefore, the pressurizer.
l 17 3
The reference to Item II.B.1 of NUREG-0737 should be corrected l
to reference II.B.1 of NUREG-0737.
l 17 4
The identifier NDT in RTNDT in the fourth line of the paragraph should be a subscript (i.e., RTmt).
18 2
TS 3/4.7.9 should be changed to CTS 3/4.7.9 for consistency.
18 2
The sentence which states: "The requirements of existing TS 3.7.4...." should be changed to "The requirements of existing CTS 3/4.7.9.."
Page 2 of 4 t
r
~
o DRAFT SAFETY EVALUATION FOR PROPOSED CONVERSION l
TO IMPROVED STANDARD TECHNICAL SPECIFICATIONS SNC REVIEW COMMENTS t
l Page Paragraph Change 18 3
TS 3/4.7.10 should be changed to CTS 3/4.7.10 for consistency.
{
l 18 4
The word "afrect" should be changed to "effect" in the sentence J
" Area temperatures will continue to be monitored and evaluated for j
their effect on equipment operability.
j 19 1
CTS 3/4.8.3.1 is only a Unit 2 requirement. The title should be revised by the addition of the following: "(Unit 2 Only)."
19 2
CTS 3/4.8.3.2 is only a Unit 2 requirement. The title should be revised by the addition of the following: "(Unit 2 Only)."
20 2
TS 3/4.9.3 should be changed to CTS 3/4.9.3 for consistency.
20 5
TS 3/4.9.7.1 should be changed to CTS 3/4.9.7.1 for consistency.
23 1
"UFSAR" should be changed to "FS AR" for consistency (two places).
23 1
Replace "...SNC's letters dated [SNC to supply dates)" with
.SNC's letter dated March 12,1998."
23 2
"UFSAR" should be changed to "FSAR" for consistency (two places).
24 1
Need to insert the proper date and Federal Register number.
24 2
The terms FQC(Z), FQW(Z), and FQ(Z) should be superscripted (C and W) and subscripted (Q) consistent with the STS.
24 3
The close quotes prior to the phrase " limit thermal power" should be replaced with the open quotes (two places).
24 4
A space should be added after the 2 symbol in the phrase "2 50%
RTP."
26 2
The sentence which begins "With a deviation alarm setpoint ofi 5 degrees F, one loop Tavg., " should be revised to state "As Trefis compared to the median signal, one loop Tavg.. "
26 3
The required steam generator level to ensure that the steam generator tubes remain covered is 28% (narrow range) in Mode 3 and 74% (wide range) in Modes 4 and 5. This paragraph should be revised to reflect this. In addition, ITS SRs 3.4.6.2 and 3.4.7.2 should be reference in the title and the body of the paragraph.
29 4
TS 3.7.1.5 should be changed to CTS 3/4.7.1.5 for consistency.
30 1
TS 3.7.1.5 should be changed to CTS 3/4.7.1.5 for consistency.
30 3, 4 The discursions ofinoperable MSIVs should indicate that the number applies on a "per line" basis to prevent potential confusion.
31 5
TS 3.7.4 should be changed to CTS 3/4.7.4 for consistency.
32 3
The reference in the title should be changed from "ITS 3.8.2.1" to "ITS SR 3.8.2.1."
32 3
CTS 4.8.1.2 should be changed to CTS SR 4.8.1.2 for consistency.
33 3
CTS 4.8.1.2 should be changed to CTS SR 4.8.1.2 for consistency.
33 5
Item (4) should be added as follows: "(4) verifying the automatic l
Page 3 of 4
1 DRAFT SAFETY EVALUATION FOR PROPOSED CONVERSION j
TO IMPROVED STANDARD TECHNICAL SPECIFICATIONS J
SNC REVIEW COMMENTS l
Page Paragraph Change i
load sequence timing capabilities of emergency load sequencers."
34 1 and 2 The phrase " subsystems are required to be operable" should be changed to " subsystems may be required to be operable." The following sentence should be added before item (1): "In the case where the requirements of LCO 3.8.10 call for ponions of a second train to be OPERABLE,". In item (1), the phrase " connected to the required direct current" should be changed to " connected to the respective direct current."
34 4
The title ofITS 3.8.10," Distribution Systems - Shutdown," should be added, consistent with the inclusion of the title for ITS 3.8.8.
34 7
The first sentence of paragraph 4 should be revised to state:
... consistent with the guidance provided in NUREG-1431.." for clarity.
34 8
Surveillance Requirement SR 3.8.1.9 should be added to the list of i
surveillances which are applicable but not required to be performed.
35 1
A description of SR 3.8.1.9 should be' added. The description of SR 3.8.1.11 should be revised to correct the omission of the loss of voltage signal stan. The description of SR 3.8.1.13 should be revised as follows: ".. 2 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at greater than or equal to a specified load."
35 4
The word " deleting" is misspelled and should be corrected.
36 1
The word " list" in the last sentence of the paragraph should be j
changed to " lists."
37 2
The last sentence of the paragraph should be revised as follows:
" Proposed ITS 3.8.9 revises the CTS LCO action statement such that with one of tbo 125-volt distribution trains inoperable, the associated service, water train is declared inoperable immediately."
(see RAI 3.8.4-01 and 3.8.9-03 responses) 37 3
The information concerning the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> restoration action should be deleted as this option is no longer in the ITS (see RAI 3.8.4-01 and 3.8.9-03 responses) 38 1
The information concerning the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> restoration action should be deleted as this option is no longer in the ITS (see RAI 3.8.4-01 and 3.8.9-03 responses) 38 4
The SE text related to the RCP Flywheel Inspection Frequency needs to be inserted. SNC needs to review the SE text to complete the review of the SER.
39 4
Need to insert the proper date and Federal Register number.
1 The tables were not provided with the draft SER. SNC needs to review the tables to complete the review of the SER.
Page 4 of 4
ATTACIIMENT II SNC MARKUP JOSEPII M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2, DRAFT SAFETY EVALUATION FOR PROPOSED CONVERSION TO TIIE IMPROVED TECIINICAL SPECIFICATIONS
pt49 k UNITED STATES g
NUCLEAR REGULATORY COMMISSION
,{
E WASHINGTON, D.C. 20665-0001 s...../
A
~ 6,m SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACT R REGULATION RELATED TO AMENDMENT NOS. A TO FA OPERATING LICENSES NPF-D NP
'y JOSEPH M. FARLEY NUCLEAR PLANT. MAND UNIT L 21Ta1 f Nimus.
SOUTHERN NUCLEAR OPERATbG COMEdN i
364
'I DOCKET NUMBERSr omA AQ l'l I i L INTRODUCTION Joseph M. Farley Nuclear Plant Unit d' Unit 2, :NP) hasSeen operat g ' h Technical Specifications (TS) issued with original operatipg@licensey" Sy letter dated Marc June 25,
, for Unit 1 and March 31,1981 for Unit 2, as amef&d from{ir tortime x
letters dated'Ekril 24,1998kAE805iI6,1998, Nove r 20,1998, supplemented bp'5[ April 3(fkO59Qfay 28,1999,Tne 30,1999, February 20h@9 July 27,199 outhern Nuclear Opera C6mpany (SNQ%icensee) proposed to amend Appendix A of Operating Licenses No. NP
"$Nd'NPF-8 to cornpktely revise the FNP TS. The proposed amendment was based upon }5,(133), " Star %A~rhi5chnical Specifications - Westinghouse Plants,"
Revision 1, dated 1995 - and upon' guidance in the "NRC Final Policy Statement on Technical SgcEcationDMYeYenfs for Nuclear Power Reactors" (Final Policy Statement),
publishedeon July 22,199 39132). The overall objective of the conversion, consistent with the fSal Policy State o rewrite, reformat, and streamline the TS for FNP to be in accordNce with 10 CFR 36.
TS are referred to as the/nproved TS (ITS), the existing FNP TS are Here er, the pr os ref "5dI5'as th ri$hi TS (CTS), and the TS in NUREG-1431 are referred to as the/andard TS e cod 5Yponding TS Bases are ITS Bases, CTS Bases, and STS Bases, respe in addition to basing its ITS on STS and the Final Policy Statement, SNC retained portions of the CTS as a basis for the ITS. Plant-specific issues, including design features, requirements, and operating practices, were discussed with thp-SNC during a series of conference calls and meetings that concluded onaprii 70.1999F in addition, SNC proposed matters of a generic nature that were not in STS. The N b staff requested that SNC submit such generic issues as a proposed change to STS through t ie Nuclear Energy Institute's Technical Specifications Task Force (TSTF). These generic issue were considered for specific applications in the FNP ITS.
Consistent with the Final Policy Sta ment, SNC proposed transferring some CTS requirements to SNC-controlled documents. In ddition, human factors principles were emphasized to add clarity to the CTS requirements ing retained in the ITS and to define more clearly the LM -Ox, -kiha.} dode d cenfett*6 CXb M Mf4 5
2-appropriate scope of the ITS. Further, significant changes were proposed the Bases to make each ITS requirement clearer and easier to understand.
The Commission's proposed action on the FNP application men 6
~
March 12,1998, was published in the FEDERAL REGIST May 8218).
The Staff's eveluation of the application and supplemental ormati t res
_ tlR requests for information and discussions with SNC during R
review is enfsd-
)
this Safety Evaluation (SE). These plant-specific change larify the ITS w s ' ct to the guidance in the Final Policy Statement and STS. There',rg frej anges are within the j
scope of the action described in the initial and supplemental.
EGISTER notices.
During its review, the NRC staff relied on the Final Pol tateme S as guidance for acceptance of CTS changes. This SE provides um basis C staff conclusion that FNP can develop ITS based on S m
d by p pecific changes, and that the use of the ITS is acceptable for contindfope taff also acknowledges that, as indicated in the Final Policy Stateme$f, the co vers s a voluntary process.
Therefore, it is acceptable that the ITS di from
,refi
- e current licensing basis for FNP. The NRC' staff approves SNC's ar)ges t CTS w modifications documented in the
~
revised submittals.
For the reasons stated infra i tilis SE, the N C ff inds that the TS issued with this license amendmentgply with Sect 6QE2a of the IdM6 iiergy Act,10 CFR 50.36, and the guidance in the Fifidi Policy'S'tEtemEt,@and th dy are in accord with the common defense and security aiEd'5$fde adequatENo ction of the health and safety of the public,
- 11. BACKGROU Section 182 f the Atomic ner requires that applicants for nuclear power plant operating licenses state:
S ' ch technical spe ations, including information of the amount, kind, and source of u
if66ility, and such oth[er information as the Commission may, by rule ETissa in.oIdD to enable it to find that the utilization... of special nuclear material will 7tf
$it'h the common defense and security and will provide adequate protection oIT!Ti and safety of the public. Such technical specifications shall be a part of any license issued.
In 10 CFR 50.36, the Commission established its regulatory requirements related to the content of TS. In doing so, the Commission placed emphasis on those matters related to the prevention of accidents and the mitigation of accident consequences; the Commission noted that applicants were expected to incorporate into their TS "those items that are directly related to maintaining the integrity of the physical barriers' designed to contain radioactivity." Statement of Consideration," Technical Specifications for Facility Licenses; Safety Analysis Reports,"
33 FR 18610 (December 17,1968). Pursuant to 10 CFR 50.36, TS are required to include items in the following five specific categories: (1) safety limits, limiting safety system settings and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance
u
)
~
requirements (SRs); (4) design features; and (5) administrative controls. Howeve rule does not,specify the particular requirements to be included in a plant's T For several years, NRC and industry representatives have s o de ines for improving the content and quality of nuclear power plant T n Febru Commission issued an interim policy statement on TS imp.ements erim Iate on Technical Specification improvements for Nuclear Po r$eact (52 FR 3 the period from 1989 to 1992, the utility Owners Groups a.thte
'C staff develo ved standard technical specifications that would establish mode 9 "Q,, mmission's policy for o
each primary reactor type. In addition, the NRC staff, licen
~snd ners Groups developed generic administrative and editorial guidelines in the fo
- a
. ' etsz 1 e" for preparing TS, which gives greater consideration to human factors p
- les and youghout'the development of licensee-specific ITS.
In September 1992, the' Commission issued G-1 i
veloped using the
{
guidance and criteria contained in the Com sion's i ri ment. STS were established as a model for developing IT[$n of t Tnterim y statement criteria to generic West nouse general. STS reflect the results of a detailed review of the applidat system functions, which were publisl[ edin's "S.jeport" ed to the Nuclear Steam System discussions concerning various'drafE of STS, s6"{tfiktSupplier (NSSS) Own ect the results of extensive i
' application of the TS criteria and the Writer's GuideMuld consi te7n$~Nflect deta EdysW51 configurations and operating characteris 11NSSS such,Te generic Bases presented in NUF.EG-1431 provide an a n82'ognfo
'arding the extent to which the STS present requirements th 'IriMeessary t blic health and safety, On July 22,1 satisfying i _g,41 e m,th s niss its Final Policy Statement, expressing the view that guidance in ytatement also satisfies Section 182a of the Act and 10 CFR
.36 (58 FR 3913 Final Policy Statement described the safety benefits of the
~ improved,STS, and encour censees to use the improved STS as the basis for plant-speci TS amendments, for complete conversions to improved STS. Further, the Final Poli.
tement gave ance for evaluating the required scope of the TS and defined the gumance enteria to b[g ed in determining which of the LCOs and associated surveillances shogr$dal in Ahed The Commission noted that, in allowing certain items to be relocated to licensehn to documents while requiring that other items be retained in the TS, it was adoptin"g th%a itative standard enunciated by the Atomic Safety and Licensing Appeal Board in Portland General Electric Co. (Trojan Nuclear Plant), ALAB-531,9 NRC 263,273 (1979).
There, the Appeal Board observed:
[T]here is neither a statutory nor a regulatory requirement that every operational detail set forth in an applicant's safety analysis report (or equivalent) be subject to a technical specification, to be included in the license as an absolute condition of operation which is legally binding upon the licensee unless and until changed with specific Commission crproval. Rather, as best we can discem it, the contemplation of both the Act and the regulations is that technical specifications are to be reserved for those matters as to which the imposition of rigid conditions or limitations upon reactor operation is deemed
4 '
4 necessary to obviate the possibility of an abnormal situation or event iving o an immediate threat to the public health and safety.
By this approach, existing LCO requirements that fall within fy an eria in the [
Final Policy Statement should be retained in the TS; those require i fall within or satisfy these criteria may be relocated to license _ ntrolled ume Commission codified the four criteria in 10 CFR 50.36 (6 36 uly 19,19 Th&
al Policy Statement criteria are as follows:
Criterion 1 Installed instrumentation that is used to detect, icate i room, a -
significant abnormal degradction of the react ressu ry.
Criterion 2-A process variable, design feature, ~operat estrict is an initial condition of a design basis accident or transientanplysis either sumes the failure of or presents a challenge to the integrity of a
' pro arrier.
Criterion 3 A structurershstem, o component that is pa of the primary success path and which functionT6Ia8tUstes to mlij Ydesign basis accident or transient that either assumes the failureP7ents a ci the integrity of a fission product barrier.
Criterion As ure sy' stem, nent which operating experience or probabilistic safety as$essmen,t has sho significant to public health and safety.
Part this SE expla e NRC staff conclusion that the conversion of the FNP CTS to those basg3kSTS, as mo by plant-specific chang'es, is consistent with the FNP current lice dtiasis and Ifeiequirements and guidance of the Final Policy Statement and 10
'50:
lil. EVALUATION The NRC staff's ITS review evaluates changes to CTS that fall into five categories defined by SNC and includes an evaluation of whether existing regulatory requirements are adequate for controlling future changes to requirements removed from the CTS and placed in SNC-controlled documents.
In addition to the initial submittal of March 12,1998, as supplemented, the NRC staff review identified the need for clarifications and additions to the submittal in order to establish an appcopriate regulatory basis for translation of CTS requirements into ITS. Each change proposed in the amendment request is identified as either a discussion of change (DOC) to CTS
m.
or a justification for deviation from STS. The'NRC staff comments were document' As requests for additionalinformation (RAls) and forwarded to SNC. SNC pro 7xied. rItten responses to the NRC staff requests in supplemental letters indicated ab8Ie'.hIie docketed letters clarified and revised SNC's basis for translating CTS rfullemenINr81Y$Qhe NRC staff finds that SNC's submittals provide sufficient detail to afow the stafgd#Efia conclusion regarding the adequacy of SNC's proposed changes.
7
- 9 The license amendment application was organized such t i a ' s were include ~
a of the fo!!owing CTS change categories, as appropriate: admi hanges, technical changes - less restrictive (specific), technical changes - le eneric), technical changes - more restrictive, and relocated specifications:
(1)
Administrative Changes, (A), i.e., non-te n es in t a sentation of existing requirements.
(2)
Technical Changes - More Restric e,(M),
new oria. onal CTS requirements.
(3)
Technical Changes - Less Res$i ve (
cific), (L),.e., changes, deletions and relaxstions of existing TS re "uku ent dss Restrictiv movei$ent of informa
~ be.? deletion of existing TS requirements by Technical Changes -)tidii $nd requirements' frdm existing specifications (that are (4) otheEE$bing rei$I lESNC-conEr'611eS documents, including TS Bases.
(5) Tech es - Less he, (LB), i.e., relaxation of existing TS requirements by providinEdiisilowance t YEsimulated or actual signal to verify the automatic actuatiorfbfYFfiIc$m nedin the surveillance test requirements of the TS.
R[ located Specifi onsj R), i.e., relaxations in which whole specifications (the LCO 6
e g
(a*nd associated actyn"a'nd SR) are removed from the existing TS (an NRC-controlled
.c document) and placed in SNC-controlled documents.
Thegejneral categ3derIof changes to SNC's CTS requirements and STS differences may be bett rgn gdpsjollows:
A. A rEtElve Changes l
Administrative (non-technical) changes are intended to incorporate human factors principles into the form and structure of the ITS so that plant operations peLonnel can use them more easily.
. These changes are editorialin nature or involve the reorganization or reformatting of CTS requirements without affecting technical content or operational restrictions. Every section of the ITS reflects this type of change. In order to ensure consistency, the NRC staff and SNC have used STS as guidance to reformat and make other administrative changes. Among the changes proposed by SNC and found acceptable by the NRC staff are:
(1)
Providing the appropriate numbers, etc., for STS bracketed information (information that must be su plied on a plant-specific basis and that ma change from plant to plant).
(O Tednniw. chwys-Les fad <d. We G.c) t.c. rem-bn of ex,3 -
T3 royivwcMs by yeovig-Y.< ey,N 6% e,in some cases mea 3
' epj g
wA% %.yteemanh o.5hAkkJ Ts, h5 pc.ecMm on oahrwkst, e L h an en4ry, s% A4 % re comanA 4 Leo 3.o.3 entry, h be con 6nuA.
wordg -tin re v-K reyhwls # to CE r, Fo.]3 anoc1%L wih suc
- l
.(2)
Identifying plant-specific wording for system names, etc.-
(3)
Changing the wording of specification titles in STS to nform toy $Ihti" ant practices.
~~(4)
Splitting up requirements cuirentlp gr'oup'ed unde ~
Ing! ~ trent speci more appropriate locations in two or more specification I
(5)
Combining related requirements currently presented i rate specifications of the CTS into a single specification of ITS.
F Table A lists the administrative changes proposedy1TSg' des'c 'ption adr' y the able A is b
corresponding ITS section DOC, and provides a guiiirrff ninistrative
[hl ff reviewed all of the change that was made, and CTS and ITS LC erence administrative and editorial changes propose by SN nd f7diis acceptable, because they j
- lI%ubstantive change in are compatible with the Writers Guide and'STS, dopo resul a
operating requirements and are consistyntYwith the.Commis '
's regulations.
~
B. Technical Changes - More Res ve Ar SNC, in electi implem rt the specifications of t
more restricfiv Ein'those(in t'IEUT$ TS reMSTS proposed a number of requirements ments in this category include requirements that are eitheMEinTre conslM5A than corresponding requirements in the CTS, or that have additional EisfrIEti8n% that afe"55 tin"the CTS but are in STS. Examples of more restrictive operable, mor,ej,est)nEtrequirements are p'l $ii{a O on'
[7quipment which is not required by the CTS to be require ent o restore inoperable equipment, and more restrictive e
SRs. Tabglists allth g_ trictive changes proposed in ITS. Table M is organized b the corresponding ITS sec and provides a summary description of the more restrictive chan e'fhat was adopted, a TS and ITS LCO reference. These changes are additional restr s on plant operatiSi that enhance safety and are acceptable.
C.
Ical Chang Less Restrictive (Specific)
Less es Ive requ rements include changes, deletions and relaxations to portions of CTS requi(re'Eelfdf5t are not being retained in ITS or relocated to a SNC-controlled document.
When requirements have been shown to give little or no safety benefit, their change or removal from the TS may be appropriate. In most cases, relaxations previously granted to individual plants on a plant-specific basis were the result of (1) generic NRC actions, (2) new staff positions that have evolved from technological advancements and operating experience, or (3) resolution of the Owners Groups comments on STS. The NRC staff reviewed generic relaxations contained in the STS and found them acceptable because they are consistent with current licensing practices and the Commission's regulations. The FNP design was also reviewed to determine if the specific design basis and licensing basis are consistent with the technical basis for the model requirements in STS, and thus provide a basis for ITS.
7 sh A significant number of changes to the CTS involved changes, a
relaxations to portions,of CTS requirements evaluated as Categories I throug at foiw:
Category 1 - Relaxation of Modes of Applicability Category II - Relaxation of Surveillance Frequency Category lli - Relaxation of Completion Time Category IV - Relaxation of Required Actions Category V - Relaxation of Surveillance Requireme Accept g
Category VI - Relaxation of LCO Category Vil - Deletion of SR Category Vill - Deletion of Requirerie for 3 y Spec eport to NRC The following discussions address p"are not r dir6d included in ITS.
1y vario S withi h of the eight categories of information cific requiremfents Relaxation of es of A litv (Category 1) 3 Reactor ope ditions CTS to define when the LCO features are required
~5Ypkigbilities%@l or power operating condition Applicab to be operable specific defined terms of reactor conditions: hot actor critica shutdown,g also be more general.
ing on the circumstances, CTS may require that th'e LCO be maintaln'id within limits i Todes" or *any operating mode." Generalized applicability condiiEns are not conta STS, therefore ITS eliminate CTS requirements such as. "all mo85s" or "any operati mode," replacing them with ITS defined modes or applicable cdnditions that are con tent with the application of the plant safety analysis assumptions for "IEbility of the re uTr'ed features.
Ingno j
. tion of this type of change, CTS requirements may be eliminated during i
conditions r which the safety function of the specified safety system is met because the feature is performing its intended safety function. De.leting applicability requirements that are indeterminate or which are inconsistent with application of accident analyses assumptions is acceptable because when LCOs cannot be met, the TS are satisfied by exiting the applicability thus taking the plant out of the conditions that require the safety system to be operable. These changes are consistent with STS and changes specified as Category I are acceptable.
Relaxation of Surveillance Freauency (Category ll)
CTS and ITS surveillance frequencies specify time interval requirements for performing surveillance requirement testing. Increasing the time interval between surveillance tests in
k 8-Ob A
the ITS results in decreased equipment unavailability due to e which iso iner ase>s
. equipment availability. In general, the STS contain test frequencies th are
, sistent with
. Industry practice'or industry standards for achieving acceptable level ment reliability.
Adopting testing practices specified in the STS is acceptab$ised ign, like-component testing for the system application and the availability of ot nts which provide regular checks to ensure limits are met.
Reduced testing can result in a safety enhancement be the unavailability reduced; in turn, reliability of the affected structure, syst nent should remain constant. Reduced testing is acceptable where operati industry practice or the industry standards such as manufacturers' recomme ions thatthese components usually pass the Surveillance wheri d at t Gnterval. Thus, the frequency is acceptable from a reliability [sta rveilla
.lency changes to incorporate alternate train testing have been
.o
.seceptable ere other qualitative or quantitative test requirements are requir.ep which Sstliftilis edictors of system performance, e.g., a 31 day air flow test an indica r t ressure in a controlled space will be maintained because this st woul(dse the me ns as the less frequent ITS 36 month pressurization test and indus ry expgrience sh
'that components usually pass the pressurization test. Additionalf('sVrveilland frequency extension can be based on staff-report analyjes bound theplant-[shecific d/digWsria component rel These chan Els are consistent %ith STS aQ6gls specified as Category ll are accepta Relaxation of om n Time
.Ill)
Upon discyo gry of a failurKogeet CO, STS specify times for completing required
, actions o me associated (l m"e$Eres that must be taken within specified comp S conditions. Required actions of the associated conditions are usedt tablish remedEa (all outage times).
Wiimes define limits during which operation in a degraded c
ition is permitted.
g completion, es from the STS is acceptable because completion times take into
.accou 't Tiiity status of the redundant systems of TS required features, the capacity Eemaining features, a reasonable time for repairs or replacement of required j
a
.ca fea Fe e low probability of a design basis accident (DBA) occurring during the repair j
period. These changes are consistent with STS and allowed outage time extensions specified as Category lli are acceptable.
Relaxation of Reouired Actions (Category IV)
CTS require that in the event specified LCOs are not met, penalty factors to reector operation, such as resetting setpoints, and power reductions shall be initiated as the method to reestablish the appropriate limits. The ITS are constructed to specify actions for conditions of required features made inoperable. Adopting ITS action requirements for exiting LCO applicabilities is acceptable because the plant remains within analyzed parameters by performance of required actions, or the actions are constructed to minimize risks associated
.g.
4 with continued operation while providing time to repair inoperable featur s. Su tions add
. margin,to safety or verify equipment status such as interlock status for f operation, thereby providing assurance that the plant is configured ap latel ions that could
- $ensurate with result in a challenge to safety systems are exited in a tim d th r
the safety importance of the system. Additionally, otherj anges to placing the reactor in a Mode where the specification no Enger a
., us i
en extension to the time period for taking the plant into Ih'utdo ditions.
are commensurate with industry standards for reductioI al power in an fashion without compromising safe operation of the pla 1ese changes are consistent with STS and changes specified as Category IV are accept F_elaxation of Surveillance Reauirement Acceptance Criteria (C CTS require safety systems to be tested and rif' ope 4 pnor enng applicable conditions. ITS provide the additional req i ment t '
by actual or test conditions. Adopting the STS allowance or "actu conditong cceptable because TS l
required features cannot distinguish be een an ual"sQrial'or a " test" signal. Category V
~also includes changes to CTS req,mpnts t are replad ed in the ITS with separate and q
distinct testing requirements whic n co ed inci operability verification of all TS.
j required components for the fea s speci1 in th
. Adopting this format preference in the STS is acceptable bec'au S SRs Wrim include testing of all previous features for survellia[nceTsting of'serified opra
' TS prov _jall6wance to bypass an inoperable channel required tp
[
annels. ITS provide the allowance to bypass the inoperabi en en ma uired setpoint adjustments on the other channels as well as performing ance test en annels. CTS test extensions allow inoperable channels to YiedfoJ;surv2iiar7:e# testing when sufficient equipment is required to be operable req EE6ts to provYe an acceptable level of safety system protection. SR relaxatio clude th
"%jtiopin ITS that administrative controls exist which provide assu nce that any chan$
omponent status, such as valve position, are recorded and tra
.' Thus, ITS exter8
' option to verify penetration integrity by administrative control to j
va s outside contain t, whereas CTS permits this option only for valves inside
- ed. nment. These nges are consistent with STS and changes specified as Category V fiIIice table.
(Category VI)
W CTS provides lists of acceptable devices that may be used to satisfy LCO requirements. The ITS reflect the STS approach to provide LCO requirements that specify the protective limit that is required to meet safety analysis assumptions for required features. The protective limits replace the lists of specific devices previously found to be acceptable to the NRC staff for meeting the LCO. The ITS changes provide the same degree of protection required by the safety analysis and provide flexibility for meeting limits without adversely affecting operations since equivalent features are required to be operable.. These changes are consistent with STS and changes specified as Category VI are acceptable.
e Deletion of SR (Category Vil)
'Both CTS and ITS include applicability requirements which
, ify th allu stitute ' {go meet an SR or failure to perform an SR within the specified time inte ilug^jo meet conver ilhCTS SRs operability requirements for an LCO. As an adjunct to tt are reviewed to establish an appropriate level of testingd LCO re.g et 7 in neme ITS. One outcome of this review is a determination tha riste to ma 5
CTS SRs. CTS SRs can be deleted as a result of ado '
rmat; such as e Mg detector testing for components that are not susceptible then simplifying surveillance test by revised testing criteria. CTS SRs c
. ' re " with a like-test that
. verifies operability of components but at a less freque. est i
- use the conditions required for testing make it safe to reduce testing si ther in vailableio ensure components are operable. CTS may also'conta n cific r ~ u ~ "ents to perform testing which verifies a criterion of a compo
. s icit co nt verification is subject to TS requirements to establish co nent 3 testing is simplified in the ITS by eliminating such narrowly foc d test r
nce frequency requirements for components may be Yised t rrespori 6: industry standards resulting in SR interval exte ions. Relaxation
. SR car made eleting the requirement to perform a SR st for a class of 6ent for com nts whose status can be adequately controlled by STS inistr "me tions add flexibility to testing where determinations to include cp'riipdri$nts in a e can be evaluated based on the status allow tesfin[g @$r?idpairs o cifEdd lant coo [d6s which as a result temporarily el of the com ent during i compon protection'aff ed the re (component. Category Vil changes include deletion or modification of illan esti 're.quirements not needed to establish equipment operability.
Deletion Recuireme -
av Soecial Reoort to NRC (Category Vill)
-CTS ude requiremen bmit Special Reports when specifiedlimits are not m.et.
Ty[tnlnates the TS ad~ y, the time peri r the report to be issued is within 30 days. However, the STS all trative contr'ol requirements for Special Reports and instead relies ei q
reporting r ments of 10 CFR 50.73. ITS changes to reporting requirements are
' ble beca 10 CFR 50.73 provides adequate reporting requirements, and the d
a s'
E, "not affect continued plant operation. Therefore, this change has no impact on operation of the plant. Additionally, deletion of TS reporting requirements reduces the administrative burden on the plant and allows efforts to be concentrated on restoring TS required limits. These changes are consistent with STS and changes specified as Catego acceptable.
Table L l~g i
requirements that have been deleted and which pertain to Category I i
through@
nd to the specific listing of changes discussed above. Table L includes all LB changes and is organized by ITS section which specifies: the section designation followed by
- the DOC identifier, e.g.,1.1 L.1 (ITS Section 1.1, DOC L.1); a summary description of the change; CTS and ITS LCO references; a reference to the specific change category as discussed above (if applicable); and a characterization of the DOC.
j we, c6 4 us c<mM dr4 & 5T5 d c%
s p fa) >9C % g m_ e wr4.W.
1 3
t 1 1.*
For the reasons presented above, these less restrictive requirements are acceptab cause they will not affect the safe operation of the plant. The TS requirements thRren-are consistent with current licensing practices, operating experien lant d transient analyses, and provide reasonable assurance that public he sa ected.
]
D. Relocated Less Restrictive Requirements
- When requirements have been shown to give little or no s fit, their remov TS may be appropriate. In most cases, relaxations previ to individual plants on a plant-specific basis were the result of (1) generic NRC act natutaff positions that have evolved from technological advancements and operating perieTesa213Mesolution of the Owners Groups comments on STS. The NRC staff re 7 genef 061amions contained in d
STS and found them acceptable because they ar lt with cur sing practices and the Commission's regulations. The FNP n.w s-ao viewed etermine if the specific design basis and licensing basis are isten
.enr.
basis for the model requirements in STS, and thus provide a ba for IT A
gi3mber of changes to the CTS involved the removal of specific re ' ments detail rmation from individual specifications evaluated to be Types gh 4 t follow:
Type 1 Details of system desig system escri t including design limits A
Type 2 De tions of e
ratio Type 3 Pr ailsf ents and related reporting problems Type 4 Admin's
'uireme dant to regulations The followipp iscussion reghy each of the four types of information or specific require is are not requi included in ITS.
Defa~ils of System Desia nd System Descriotion includina Desian Limits (Type'1) l esign of the f is required to be described in the SAR by 10 CFR 50.34. In additkth,e,quji. assurance (QA) requiremergs of Appendix B to 10 CFR Part 50 require l
t t' design documented in controlled rocedures drawings, and maintained in l
IdE$lth an NRC-approved QA plan ((lE ac SAR Ch er 17). In 10 CFR 50.59 controls l
are specified for changing the facility as described in the SAR, and in 10 CFR 50.54(a) l criteria are specifie for changing the QA plan. In ITS, the Bases also contain descriptions of l
system design. I 5.5.14 specifies controls for changing the Bases. Removing details of system design fr the CTS is acceptable because this information will be adequately controlled in the SAR, controlled design documents and drawings, or the TS Bases, as appropriate. Cycle-specific design limits are moved from the CTS to the Core Operating Limits Report (COLR)in accordance with Generic Letter (GL) 88-16. ITS Administrative Controls are revised to iriclude the programmatic requirements for the COLR.
p.
Descriotions of Systems Ooeration (Type 2) f
'The pibs for the normal and emergency operation'of the fag lity are be described
. in the VFSAR by 10 CFR 50.34. ITS 5.4.1.a requires wrig piroced Sgstablished,
' implemented, and maintained for plant operating proce s includin
.G r ommended in Regulatory Guide (RG) 1.33, Revisior lAppen
, eb l
~
ntrols specified in 10 CFR~ 50.59 apply to changes lj ocedi as descril SAR. In ITS, the Bases al contain descriptions of ration. It is aGiinte,Ule to remove details of system o ation from the TS becaus f information will be adequately controlled in the SAR, plant operating pr he TS Bases, as j
appropriate.
Procedural Details for Meetina TS ReauirementsMed Report muu ems (Type 3)
)
boecdd the plant Details for performing action and SRs e app R,an S Tal iEFexampie,controiof the procedures required by ITS 5.4.1, the
.4 plant conditions appropriate to perfo survel test brF Esue for procedures and
-scheduling and has previously be prmin be un ssary as a TS restriction. As indicated in GL g1-04, allowing th du ontroli sistent with the vast majority of other SRs that do not dictate pl onditi surve nces. Prescriptive procedural information i an action regufrergent is un!
sin all procedural considerations g
ra o compi
.ac ions required, and referral to plant necessargfor the plant proceduresTrefore requ
_y eve her changes to procedural details include those ass
.Jmits he ITS. For example, the ITS requirement may refer to programma uirements as LR, included in ITS Section 5.5, which specifies j
the scope of t edi d, LR and mandates NRC approval of the analytical l
methodol l
l The val of these ki c
ral details from the CTS is acceptable because they will dequately contr he FSAR,' plant procedures, Bases and COLR, as a
riate. This appr h provides an effective level of regulatory control and provides for
~
' appropriate c ge control process. Similarly, removal of reporting requirements f
Os is ap e because ITS 5.6,10 CFR 50.36 and 10 CFR 50.73 adequately co med to be necessary.
l Ad Reauirements Redundant to Reaulations (Type 4)
Certain CTS inistrative requirements are redundant to regulations and thus are relocated to t or other appropriate SNC-controlled document The Final Policy Statement allows licensees to relocate to licensee-controlled documen CTS requirements that do not L
meet any of the criteria for rnandato c usion in the TS.
hanges to the facility or to procedures as described in th SA are made in accor nce with 10 CFR 50.59.
hanges made in accordance ' ith the provisions of oth r licensee-controlled documents are g
w subject to the specific requirements of those docume
. For example,10 CFR 50.54(a) govems changes to the QA plan, and ITS 5.5.1 gov rns changes to the Offsite Dose Calculation Manual (ODCM). Therefore, relocati of the administrative details identified above, is^ acceptable, o r d ele O $ c.e n M0d th Accumed Such as, %e C.de A FeAcal %ob dmi. er RaybAwy Gv'.Au 4
. ih c.W6rdcmorA6kok Table LA lists CTS specifica"ons and detailed inf f ati removed from i viduals cifications that are relocated to SNC-controlled documents T. Table LAis orga
. TS section and includes the following:
p.
DOC identifiers, e.g., Section 2.0,11-LA (ITS Section.
DOC 1 nificant Hazards Evaluation")
CTS reference the name of the document that retains the CTS req a
the summary description of the change the method for controlling future changes to relocatedfequ reme a reference to the specific change type, as discuss abo jocluding the u
information or specific requirements in ITS The NRC staff has concluded that these types of iS stalledirl orm specific requirements f
are not necessary to ensure the effectivenes dfl t
the health and safety of the public. Accordingly, these requireme
'may b controlled docume is for wh a ea, u
ove englatory or TS RAenT Re3og, s
mea,
ce requirement: qW Reol b.
G,
' or er (1)T Bases controlled y ITS 5%14 "Tekni' cal Spepif$tions Bases Control Program."
~
10 CFf34(includes the Tectinich[Requiref6*E6t1 Man,ual (TRM) by SAR (2
(3)The ODCM,jg.
0.59@.
6 F
S co olle CFR (4)The Q is approve C and contained in FSAR Chapter 17 and controlled '
.CERPart 5 11 B.
For each of thje,se e ang LT e LA lists SNC-controlled documents and the TS or regulatory requirements goveming changes to those documents.
To the,eMent that requirem ntsend information have been relocated to SNC-controlled docuhts, such informa and requirements are not required to obviate the possibility of an i
abnormal situation or e giving rise to an immediate threat to the public health and safety.
FurtRE7where such infoJmation and requirements are contained in LCOs and associated o
requii$rfi5Ets in thegS, the NRC staff has concluded that they do not fall within any of the four criteyi,[n]th(Fin [a) Policy Statement (discussed in Part 11 of this SE). Accordingly, existing detaile~d infonhabon and specific requirements, such as generally described above, may be deleted from the CTS.
E. Relocated Specifications The Final Policy Statement states that LCOs and associated requirements that do not satisfy or fall within any of the four specified criteria may be relocated from existing TS (an NRC-controlled
~
document) to appropriate licensee-co olled documents. These requirements include the LCOs, Action Statements (ACTIONS) and associated SRs. In its application, SNC proposed relocating such specifications to the SAR (includes the TRM by reference). The staff has reviewed SNC's submittals, and finds that relocation of these requirements to the FSAR (and TRM) is acceptable, in that changes to these documents will be adequately controlled by (O covs APA M Fdf*D*) diO Venf; r#h l~rNr 'Tisk3
(.O T.sc Profwm
.& 6/FTP) en N hsel ceca wWnmed to spe c% ad sb e sk-pmq ub 3 a&
l Tinch kenhe b>
10 CFR 50.5g. These provisions will continue to be implemented by appr riate procedures (i.e., operating procedures, maintenance procedures, surveill
,esting procedures, and work control procedures).
SNC, in electing to implement the specifications of STS, al reposed,y ith the criteria in the Final Policy _ Statement, to entirely remove ce n TS fr y e C in SNC-controlled documents noted in Table R. Table R I L all s fications a c ficCTiS details that are relocated, based on the Final Policy State j C-controlled docupnirits in ITS. Table R provides: a DOC identification number referehTedKClS; a CTS reference; a summary description of the requirement; the name of the pug #niths retains the CTS requirements; and the method for controlling future chan a to re'iGCdb:iulrements. The NRC staff evaluation of each relocated specification cific C
~
~
sented in Table R is provided below.
Boration CTS Requirements to maintain a source of bor d wat ne or w paths to inject this 1
borated water into the reactor coolan m (R
, and a priate charging pumps to
~")
provide the necessary charging he etc reactor ssure for boron injection are relocated to the TRM; The relocati' f the fo in specifications addressing the boration subsygtem of the chenfica' and volumTEEn stem (CVCS) are addressed as a group becaufgch represg"$lement kl$r'ation subsystem and as such there are common functional requireme
- ss 11 similar relationships to DBAs:
- CTS 3/4.1.
th - S CTS 3/4.1.
-0
.2.4 Cha
- Operating
.-C
.1.2.5 Borate ources - Shutdown--
3/4.1.2.6 Borate er Sources - Operating ability of th6se tiori subsystems or' components required to mitigate a DBA or The
'are retaine Chapter 3.5 TS for emergency core cooling systems (ECCS) since they l
tra med NRC policy statement.
The system of the CVCS provides the CVCS capability to control the chemical neutron absorber (boron) concentration in the RCS and to help maintain the shutdown margin (SDM). The operation limits retained in SDM and rod insertion TS provide adequate assurance that the required parameters (SDM and rod position) are maintained within the design and
- analyses limits. The boration subsystem is not specifically assumed to be operable or credited in the applicable safety analysis to mitigate the consequences of a DBA or transient, including the limiting case of a boron dilution event. In the case of a malfunction of the CVCS causing a boron dilution event, operator response is to close the appropriate valves in the reactor water makeup system. The calculations supporting the analyses of the boron dilution event show adequate time exists for operator action to mitigate the event before SDM is lost, criticality is reached or for some form of boration to be initiated to restore the SDM (FSAR 15.2.4).
CTS 3/4.3.3.7 - Hiah Enerav Line Break Isolation Sensors Requirements for instrumentation used to either detect and m ate the of steam or water into plant areas or to provide control room operators '
rmst of a line break event are relocated to the TRM. The instrumentatiorg nsists of ssu les whic
- monitor the air pressure in piping penetration and equipmers frooms c3 inin
~
Argy la outside of containment. In addition, CTS 3/4.3.3.7 contai quirenWnts for levs
" ~
tsed to detect flooding in the main steam (MS) valve room. Th Et'rurnentation address UTS 3/4.3.3.7 functions to actuate the following isolation valves al._
toom air pressure signal:
Instrument air supply isolation valves, Nitrogen supply isol
'vs steam generator (SG) blowdown sample isolation valves, Pressurizer steam s sa '
' n valves, Pressurizer liquid sample isolation valves, RCS loops ar[d 3 som valves, EVCS letdown isolation valves, and SG blowdown isolati a
The lev es are used to detect flooding in the MS valve room and functi.
i feedw pumps.
The instrumentation addressed by CTS 3/4. 7 init i s t actugtion of equipment required to tems a 'strTitures outside containment.
The functioEs performed by the instrumen ationprevent damage to the surround TS 3/4 4 7 are not functions that are required by a safety analysis to miti a~fitne co uence 5, a design basis (line break) accident described in the FSAR. V isolatign actuat required to mitigate the consequences of design basis,
~ptureagEdEn scribed in the FSAR are performed by ESFAS signalsTdch as Ph 4 B contaignl $lation, steam line isolation, and main
- feedwater is$5t
,ese SYXiff(olation functions continue to be required operable by the ITS ESFAS L dition, t ' fee'dw,ater pump trip actuated by the flood detection instrumentation in
.3.3.7 isviirFaTed to' meeting the acceptance criteria for feedwater
^
$$(ain's adequately covered). The reactor trip signals a SA line breaks descr ensuring the jr injection acidation signal
__ hdescribed in the FSAR (15.4.2.2.1) as providing protection against in feedwater p
~~
~ 're continue to be required operable by the ITS ESFAS and RTS CTS 3 3.2 - Movabl ncore Detectors Re me s.
vable incore Detector Specification, CTS 3/4.3.3.2, used to ensure ope trumentation for monitoring flux distribution within the core are relocated to the T movable incore detector instrumentation provides information used for periodic surveillances of the reactor power distribution, and for calibration of the excore detectors. The power distribution surveillances verify that peaking factors are within the assumptions of the design analysis. These TS more directly address the required reactor power distribution limits
, and peaking factors in ITS sec' ion 3.2, Power Distribution Limits. In addition, the Reactor Trip System (RTS), LCO 3.3.1 contains SRs that require calibration of the excore detectors. As such, the RTS specification requires the movable incore detectors to support test verification that the excore detectors are oroperly calibrated by th formation provided by the incore detectors. However, the movable incore detectors do not directly address the required parameters or the equipment calibrated.
l CTS 3/4 3.4 - Turbine ersoeed Protection Requirements for existing
/4.3.4 conditions, ras, and S 4 '. the t negers ed protection system instrumentation are relocated to the TR 3 is speci r u es the
- (Ora turbine overspeed protection instrumentation and turbine ied contg, alves 1
r protect the turbine from excessive speed which prevents t ner s,ti5n of poten ~
missiles. Although the design basis accidents and transie id e a variety of syi d,Nilures
~
and conditions which might result from turbine missiles stri v
- equipment, the system failures and plant conditions could
' plant systems and her events as well as turbine failures.
The operation of the turbine overspeed protection pg tassu credited by any design basis accident analysis. A turbine missil bili sis w rformed to determine the potential of turbine missiles to
' genera upe the fai re of safety related equipment. The probability analysis, s base on nce ffrequi s of the turbine overspeed protection testing pr r rh descr'
~d in F ionho.2.2 cause CTS 3/4.3.4 does not contain SRs for
, rbine rspee otection system.
nalysis s bei gene at om turbine overspeer The showed a low likelihood of turbine,Te test relocation of this TS does not im act t
'm bed in FSAR sectiond
.2.2 and as such, the relocation of CTS,
.I$does im anaivsis of the probabiity of tu ine missiles to be $herated, s d cause t Ee"6f safety'related equipment. In view of the lowlikelihoc o iurbine overs' peed scenario does not constitute a part of the primary (su [I6r6ine miss pith to pre NtYrh te such design basis accidents and transients.
CTS 3/4.4.2 - Safe <v alves-Shut odes 4 and 5)
V Requirements ' iafety ve utdown (Modes 4 and 5) are relocated to the TRM. The pressur' fsaf, % valves p RCS from being pressurized above the RCS pressure Safet it. In WS, the pr tssu zer safety valves are required operable to provide overpressure prote from operating ditions (Modes 1-3) down to the RCS temperature at which the low T
temgEature overpressu rotection system (RHR relief valves) are required operable (Mode 4 s3;MTherefore, LCO 3.4.10, Pressurizer Safety Valves, and ITS LCO 3.4.12, Low TeriferatJreOv ressure Protection System, requirements provide continuous RCS
~
ove~
- n. As such, the CTS 3/4.4.2, Safety Valve - Shutdown, requirement for a singl su er safety valve to be operable during all of Modes 4 and 5 is not required for RCS overpressure protection. In addition, the operability of a single safety valve in Modes 4 and 5 is not an assumption of any safety analysis for the mitigation of a design basis accident or' i
transient in Modes 4 and 5.
CTS 3/4.4.8 - RCS Chemistry The RCS chemistry limits of existing
/4.4.8 are relocated to the TRM. The reactor coolant chemistry program provides limits on particular chemical properties of the primary coolant, and surveillance practices to monitor those properties, to ensure that degradation of the reactor coolant pressure boundary is not exacerbated by poor chemistry conditions. However, degradation of the reactor coolant pressure boundary is a long-term process, and there are
p
- other, direct means to monitor and correct the degradation of the reactor tant ure boundary ivhich are controlled by regulations and TS; for example, in-se ion and primary coolant leakage limits are provided to prevent long-te~
grad
_e reactor
- coolant pressure boundary materials, and provide long term nan le structural conditions of the system. These limitations are immedi to the
(
operator, and are not required to ensure operability of the S pres u
CTS 3/4.4.10.2 - Pressurizer Pressurizer temperature li,mits are relocated to the TRM. tj
_ d on pressurizer operation to prevent non-ductile failure of piping. These l@ s arTJtinsjy qct with the accepted structural analysis. Since the' pressurizer normally o in tem es above those for which there is a reason for concem of nonduct perat are placed on the enalysi rmed in pressurizer to assure compatibility of operation seJithit trendisent operating 7
accordance with the ASME Code requiremen,tapW restrictions and Criterion 260es not apply smceg s' ' rist yestrictions, Criterion 2 applies only to those operatigfrestrictionsYequired preclu juna styzed accidents and transientsh iel"ded n IafHowevfe a'fpilure fressurizg integrity would result in an fanalyzea event (loss o" coolant accid 6ntTfor w 1. numerodi ' systems and components are J
required _and retained in the TS.JT efore,t ressuriz rtemperature limits are not relied on fo prevent or mitigate a DBA or ent, nor its an operating restriction that is required to pre'cidde an un cident piiEt.
CTS 3/4.4.12 - en or Vess nts Requirements for rel ted to the TRM. The RCS Vents exhaust non-condensable gpes I rom the RCS which may inhibit natural circulation core cooling folg ng any eve a loss of offsite power and for which long term cooling, l
such asgioss-of-coolant a OCA) is required. The functional capabilities, a'nd testi req'uirements for reactor ve ad vents are consistent with the requirements of item II.B NUR 737, "Clarificati f TMI Action Plan Requirements," however, the operation of RCS Ven ot assumed in safety analysis since operation of the vents is not part of the pri uccess path ny design basis event. The operation of these vents is an operator acti
_ emve as occurred, and is only required when there is indication that natural circ urring.
CTS 3/4.7.2 - Steam Generator Pressure / Temperature Limitation Eb Requirements for the steam generator pressure / temperature limits in TS 3/4.7.2 are relocated 4
ressure induced stresses to the TRM. These pressure and temperature limits ensure that the are within the maximum allowable fracture toughness stress limits.j val of the pressure and temperature limits are based on maintaining steam generatorF
. t a level sufficient enough to prevent brittle fracture. However, if failure of steam generator integrity occurs, the plant condition that results is bounded by the analysis of a steam generator tube rupture or other loss of coolant accident events for which adequate mitigation systems and components are provided. The systems and components provided to mitigate analyzed events resulting from a failure of steam generator integrity are retained in the iTS. The Steam Generator
, Pressure / Temperature Limitation is not an initial condition of an DBA or transient, Nthis limitation an operating restriction that is required to preclude an unanalyze@ ace or transient.
/4.7.9, " Snubbers," states that all snubbers s be
'~
passive devices that are designed to prevent unrestraine n undt r dyna and allow normal thermal expansion of piping and nozzles to eli cessiv thermalstr sses during heatup or cooldown. The TS action statement for s
'n qui s that an o'
inoperable snubber be replaced or repaired. The SRs fo b
.De ri examined under the inservice inspection program.
uiremen o stirp at all snubbers be operable are requirements that do no.i.
actor o o not identify a parameter that is an initial condition assumption. la
- nsient,
'.ot identify a significant abnormal degradation of the react olant ~ essf ~
ry, and do not form part of the primary success path which functiongor actuat to ip esign basis accident or transient. Requirements for snubber o ra ality ar ocate RM.
- CTS 3/4.7.10 - Sealed Source Cont tion j
3/4.7,10, " Sealed,Sourp$e Contarga o'rfgequires that sealed s
Existin radioactive mafdiial shall be fr$tiof d specified reEvable contamination; thereby ensuring that leakage frorkIQOuct, soYcia'r7 cial nu'l!fe5 material sources will not exceed allowable values speciffe'd$10 CF,*R Pa associated action statement requires that if the removable contarhEa'tiEn exceeds *
, the sealed source shall be either decontaminated or disposed of. Tj[lirEti'tIo*ns expres$iIithis TS do not impact reactor operation or the safety of reacpopera
'rements specified in the existing TS have been relocated to the TRM.
CTS 3
.13 - Area Tem Monitorina (Unit 2 Ontv)
Requ ents in CTS 3/
3 for area temperature monitoring are relocated to the TRM.
~
TheE its ensure,that safety-related equipment will not be subjected to temperatures in exc g'r e.nvigr nmental qualification temperatures. Exposure to excessive temperatures coul rade'e ipment over the long-term and may impact equipment operability. However, equipr nt temperatures do not give a direct status of the operability of specified equipment, rather it is only one of many factors used in the evaluation of operability of safety related equipment. Ultimately the operability of safety-related equipment is determined and controlled within the TS by the definition of operability and the individual TS which require the safety-related equipment rable. Area temperatures will continue to be monitored and evaluated for their n equipment operability, in accordance with the requirements retained in the TS for the affec equipment. Therefore, the existing safety-related equipment TS and the TS definition of o ability provide adequate assurance that safety-related equipment is operable.
O CTS 3/4 8.3.1 -Containment Penetration Conductor Overcurrent Protective Devi V,M 2 Odd Requirements for CTS 3/4.8.3.1, Containment Penetration Co ctor r.
Protective i
Devices are relocated to the TRM. This LCO contains requ ents fo rcurrent I
devices and breaker position or fuse status to minimize th ntial for it in a
_.;compon_ent inside containment or within the cabling penet fon of th tain Containment electrical penetration conductors are protect d
gizing cir breaker trip or removed fuse for circuits not required duri iiBt peration or by i45llilE
. overcurrent circuit breakers which are periodically surveille operability. De-energizing an AC circuit minimizes the potential for an ele la component inside containment from propagating to equipment outside cont men ei lly damagi,ng the penetration. However,10 CFR Part 50 Appendix J l testin required monitoring of all containment penetrations for de se de ide protection for the circuit conductors against damage or failur i heat ffects, but are not relied upon in any design basis accident or tr
_ CTS 3/4.8.3.2 - Motor-Operated Valves 7Fermal riosd n Device M 2 Ordd 3
Requirements of CTS 3/4.8.3.2, Mot rate Valve The i Overload Protection Devices are relocated to the TRM. This LC ontains uireme that ensure the thermal overload protection will not prevent a safet ated m r-operI58 valve from performing its intended
,se devices {'arergtec't the eq%hrif ffom potential damage to maint safety functio capability
- ment, relied upon in the primary success path to mitigate a design basis a transie Ives protected by thermal overload devices are required to be opera ppo le opIAIsbility of their associated system. Thus, the those systempc[t erm"T8@rload protecTt n devices is sufficiently operability of the a
ont7tii%
s de%ned with such devices. Moving these requirements outside T will not, by itse the existing operability requirements for safety-related motor o ted valves or r associated operational restrictions imposed by the applicable syste Os.
CTS 9.3 - Decav T're F
Th 'CT.
D me is relocated to the TRM. The Decay Time TS requirement ~ places a time m' ongea uberiticality prior to the movement of irradiated fuel assemblies out of the reacto Eskel'.%.is ensures that sufficient time has elapsed for the radioactive decay of short-lived fission products and is consistent with the assumptions used in the fuel handling accident safety analysis.
However, the schedule restraints associated with a normal refueling shutdown always ensures the movement of irradiated fuel does not occur prior to the CTS Decay Time TS limit of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. Refueling outage schedule restraints include RCS cooldown, depressurization, boration, removal of the reactor vessel head and upper intemals, flooding the reactor cavity to the required level, as well as various required testing and maintenance activities. The activities and requirements associated with a normal refueling shutdown are inherent in the design and operating restrictions (i.e., cooldown and depressurization TS requirements, water level TS 1
~.
requirements, boron concentration TS requirements, and TS requirements o main
'and test
- eguipment to ensure operability) associated with pressurized water r act C
Although Criterion 2 of the Final Policy Statement would req
. existing af. ht" Decay Time" to be retained in the ITS, the requirement for a 100 decay tirp. ol
'subcritical before commencing movement of irradiated fuelin the reaq r vessel i. alwa >
refueling outage.-The operations required prior to moving diat
~
I in the re.
require in excess of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to complete before irradiatd be moved. T be requirement is unnecessary and has been relocated from t tions to the TRM.
CTS 3/4.9.5 - Communications Requirements contained in CTS 3/4.g.5, "Commuri propo relocated to the
~ TRM. This specification establishes requireme mun' n between the control room and the refueling station during eling that refueling personnel can be promptly informed of sig nt ch tatus or core reactivity conditions observed on the control room i rumen
- n. Thi rement represents good operational practice and is designed t[nqure s fuelin rations; however the refueling system design accident or transient pohse a not ta edit for communications.
Therefore, the requirements have rel CTS 3/4 9.6 anioufator rane:
Requireme g3/g9 Crane" are relocated to the TRM. The requirements of this specificat o sugthat h or crane and auxiliary hoist will have sufficient load Tut off limit for the manipulator crane, and load handling capacit e ove indicators for '
auxi a uring r eling operations. Additionally, this specification ensures t he core int p ' actor vessel are protected from excessive lifting force
~
during r ing ope' rations vent they are inadvertently engaged during lifting operations.
Altho
. is specification s requirements designed to prevent damage to fuel
'es, core internals d reactor vessel, these requirements are not relied upon to asse pre r mitigate the
/ sequences of a DBA.- The limitations of this specification only apply to quireme Sesign controi requirements are adequately governed by regulation Cr$
CTS Crane Travel-Soent Fu I Storace Pool Buildino Bridae Crane The crane travellimits specified in existing 4.g.7.1 support the safety analysis assumption that loads in excess of 3000 pounds will not be moved over fuel assemblies stored in the spent fuel racks. The action statements of are provided by physical design and administrative
' controls. The applicable safety analysis assumes a 3,000 pound load dropped at a height of 42 inches. The results of this analysis show that the accident would not result in damage to the fuel assemblies or an unsafe geometric spacing of the fuel assemblies. This analysis conservatively bounds the drop of a standard fuel assembly with a control rod and handling fixture at the P
'x.
maximum lift height of 39.5 inches. The only other, heavy load handled by brid aneis the spent fuel pool transfer slot gate. The transfer slot gate is moved fro I position' directly to its stored position without moving over stored fuel, ver,j racks are designed to withstand a transfer slot gate drop /There are n heav are handled by this crane.
The spent fuel storage pool building bridge crane travel li callimits accordance with the requirements provided in the applica rocedure and notegkhliss variables which are monitored and controlled by the operat
, the requirements have been relocated to the TRM.
~
CTS 3/4 9.7.2 - Soent Fuel Cask Crane -
Requirements for CTS 3/4.9.7.2, " Spent Fuel
.r
' relocate he TRM. The spent fuel cask crane is used for handling shippin scon nt assemblies. The crane transfers the shipping cask between the o trans d(at sk wash and storage
~
areas. Fuel assemblies are placed in t k by ridge
_ The cask crane _is.
prevented from moving above or into inity speni i pool by rail stops and mechanical bumpers which are per lya ed. T
. sk crane does not move loads over the spent fuel pool.
Although this. Acification gn sp,ent fue cash,re,quireme{.
ned to ensure correct operation and maintenancs uences of(a D$The limitations of this specification only a crape, these requirements are not relied upon to prevent or mitigate the c ns requirements.
n E~ontrol reqN 6feYare adequately governed by regulation and required QA plan.
CTS 3/4 9.1 Storsoe lation (Fuel Storace) t _
Requi nts of CTS 3/4.
torage Pool Ventilation (fuel storage)" are relocated to the TRM is specification co ains requirements for operability of the penetration room filtration syste enever irradia [d fuel is stored in the spent fuel pool.
ne ionJoorgjiltration system limits releases of radioisotopes to the environment which may
. CCS pump rooms and penetration rooms during the recirculation phase of a desig accident. The system also processes the air from the fuel handling area following a fuel handling accident. The system is manually aligned by operators to operate in one of two modes (fuel handling or LOCA). Normally the system is aligned to automatically process the exhaust air from the spent fuel pool area upon receipt of an actuation signal initiated by either high radiation or low flow in the spent fuel pool exhaust system. In the event of a LOCA, the system is manually realigned to operate in the LOCA mode prior to the end of the injection phase of a LOCA.
, 4 Operation of the penetration room filtration system is assumed in the safet analys' - or a fuel handling accident inside the feel handling building. Therefore, this syste st
' operable and aligned for this purpose wnen the potential for a fuel han accid
. In the refueling section of the CTS, this system is addressed in tw parate L A12 for fuel storage and 3/4.9.13 for fuel movement and crane operati ince the el handling accident exists during irradiated fuel movement t TS 3/4 l (fu
~ ~ retained consistent with the STS in the~ plant systems cha th ' -
.~Howeve
' ~-
3/4.9.12 (fuel storage) requires the penetration room filtra tem operable and "to the fuel handling building whenever irradiated fuelis storec t fuel pit. As such, the -
potential for a design basis fuel handling accident in the fue _
a only exist during movement of irradiated fuel and this is adequately addre d byWreidry id CTS 3/4.9.13.
Therefore, the CTS 3/4.9.12 (fuel storage), applicabi n there is1Muut hovement,is not relied on to prevent or mitigate the consequences asis a gend is proposed for relocation.
CTS 3/4.10.5 - Position Indication System butd
- The test exception for Position Indicatio stem hutdo ws the~ CTS 3/4.1.3.3
~~~
requirement that a single digital rod "Iti6n indi or be o
- rable for each rod not fully inserted to be suspended in Modes 3,4, an Tor the e
drop time measurements.
The control ro sition indica stem pro ation of rod position to the operator.
This indicati$ dis li' sed by thE &
verify Ifa the rods are correctly positioned. In
~
operating Mo'd5 "$$Qthis is used during reactor startup and operation to monitor rod position to ve insertion and nme limits are met and to verify that the rods are fully tely f o[w reactor trip. Rod position indication requirements inserted into the mm during startu n opera addressed in the ITS by the LCO, ' Rod Position Indication" which satis Criterion Policy Statement (verification ofinitial conditions of a DBA).
The 3/4.1.3.3 requi nt for rod position indication during shutdown, Modes 3,4, and 5, with reactor trip bre closed, specifies that a single digital rod position indicator is req be operabl r each rod not fully inserted. The associated CTS test exception
(
lowsgrequirement to be suspended in Modes 3,4, and 5 for the purpose of rod dro eas rements. in the shutdown Modes, the position indicating system provides inform nd is not relied on by the operators to verify insertion or alignment limits (which i
are only required in Modes 1 and 2). Therefore, in the shutdown Modes the rod position
)
indication system is not used to verify the initial conditions of a DBA. Additionally, during shutdown Modes, the rod position indication system is not used to verify a reactor trip, or assist in the mitigation of any other DBA or transient.
' Conclusion The relocated CTS discussed above are not required to be in t TS un 1.1 R 50.36 and obviate the possibility that an abnormal situation or event wild yE god to do not meet any of the four criteria in the Final Policy Staterr 3 ve rise t nintf id Fathreat t the public health and safety. In addition, the NRC staff fin
'that su t regat exist under the regulations cited above to maintain the efft th isions in specifications. The NRC staff has concluded that appropr _. _ s have been e for all of the current cifications, information, and requi are being moved to SNC-controlled documen This is the subject of a license herewith. Until incorporated in the SAR and procedures, changes to se s
, information, and requirements will be controlled in accordance with theja plicable c 1
ures tha't control these documents. Following implementation, the NBriwbit the re rovisions to ensure that an appropriate level of control has
~
fv geNR aff has concluded that, in accordance with the Final Policy Sta t, s
.rego to,
ntrols exist under the rdulations, particularly.10 CFR 50.59. A ingly,'
se cif ns, information, and Nuirements, as described in detail in
,ma reloca m CTA andplaced in the f SAR other SNC-controlled doc as s cified in S
's letterFdateantsNu to suppID h$k F. Control of Specification u remen
'nd1n ormation Rernoved from the C S
%w.
f The feJlity gprocedures ser
'in the i and TRM, incorporated into the FSAR by ensures records}n(MrevisedMj^ Dab ishes appropriate control over require reference, can-o arice with the provisions of 10 CFR 50.59, which a
ed a es from CTS and over<
es to-rYquirements. Other licensee-controlled documents contain provi nges onsistent with other applicable regulatory requirements:
for.examp ODCM nged in accordance with ITS 5.5.1, the emergency plan.
implem g procedures (f in be changed in accordance with 10 CFR 50.54(q); and the
' admi '
ive instructions t lement the QA Plan can be changed in accordance with 10
~
CFR (a) and 10 CF art 50, Appendix B. Temporary procedure changes are also con' by 10 CFR 5 1(a). The documentation of these changes will be maintained by Iccordance w] lations as 10 CFR 50.59.
e record retention requirements specified in SNC's QA plan for FNP SN ica gu an The li n ion for the relocation of requirements from the CTS will address the implementation of the ITS conversion, and when the relocation of the CTS requirements into licensee-controlled documents will be completed. The submittal of the updated licensee-controlled documents (e.g., FSAR) to the Commission will be as required by, and in accordance with, the regulations (e.g.,10 CFR 50.71(e) for the updated FSAR), and not be as part of the implementation of the ITS.
. G. Evaluation of Other TS Changes included in the Application for Conversio_n
%Q Irnproved Technical Specifications 4
es in which Fijfproposedr,hj{ngdto both t This section addresses the beyon -scope is of co eration for these befon -scoi$ issues in CTS and STS. The NRC publish d notice the FEDERAL REGISTER on ae 64 FR11
. The ch"a'Iges disgEsed bdiBM]isted *,
the order of the appilcable ITS specification or section, as ppropriaW.
[
ITS 3.2.1 - DeletidWOC(Z1 and EOW(ZB[ E"(7.
b (2.)
Q Q
p SNC deleted STS terr $1f0C d
W u
d ti e dge the tragn'f} steady state" limit, t
it ih it may be and ' transient" limit. Since the term
(
_s deleted.
hermore, if SNC deletes the term EDW ti le is no neeij t pecify a term
(
hes_e_ limitpy then be referred to asjtead Dand tran limits instead of TeTnic nt of the CTS QW Z. Since these changes [ddiot alte and o
nge is acce able.
requirements, the Fj(2. er Rath[r[an Re
~ ITS 3.2.4 - Limit Therm ce Therma Power W
h When the quadrant power tilt ratio jQfTR) exceeds 1.0,ihe STS also requires red e plant enters CTS action a.2.b which requires perators to rejuc% thermal p%IEVs[y 46 replaced N miijher. mal poweMI6'uring starttTf; fis phrase " reduce thermal p rmal power nder thesegdit' ions, in theQS 0PTR may exceed 1.02 because of transient core condshons cue to, xenon. Inese conditions are
!! helf-correcting as a result of power ascension and xYihth up. ChYhikthlanguage to imit thermal power" rather than
" reduce thermal powerawould allow increases in power within the mode of applicability (when j
OPTR excee [harge is accep able.02}?tIQo 9 still coEtTiue to limit power b Therefore, th,ds sc ITS 3.2l4 *- AFD und QPTR"ADollcability prd b SNC,M R
revised CTS 3/4.2.4to be consistent with the Mode of Applicability in the AFD TS, and revis%6e OPTR Apgli[Mility from "above 50% RTP" to S 0% RTP." Although this change introduIis'a slig,ht,1y1nore conservative Applicability for the OPTR LCO, it maintains the CTS cons"i!s'ib~ncTy b^et~ ween the AFD and QPTR Applicabilities. This change has no practicalimpact on pla'ntTcipe7a'tI5 and'no impact upon safety, yet it eliminates subtle differences between the LCOs. This change is acceptable.
ITS 3.3.2 - Main Steam isolation Instrumentation SNC in their ITS conversion are not adopting STS footnote (i) and are replacing it with ESFAS Table 3.3.21 footnote (d) to take credit for the FNP MS system design. FNP's MS isolation instrumentation is required to be operable during Modes 1 and 2. STS footnote (i) provides an exception to the Mode applicability for the MS line isolation function when all the MS line l
-2 5 -
1 solation valves are closed. Once the MS isolation valves are closed, the ty n of the MS isolation instrumentation is accomplished. STS footnote (i) was writt ndard MS system design which contains only one MS isolation valve per line.
otnote (") to Table 3.3-3 also requires all MS isolation valves to be closec
- ever, S system has two isolation valves in each MS line. SNC has propos change
[ ell) to reflect the FNP design since closing one isolat'on valve in h MS li g. comy
~
intended safety function. Footnote (d) to the FNP ITS Se 3.3 11 allow t the Mode applicability for the MS isolation function when,
an valve in eac
'si closed. The staff finds the proposed change acceptable.
ITS 3.3.4 - Source Renee Neutron Flux Monitor for Re stem CTS Table 3.3-9 does not include the source rang <M SRM) ote Shutdown System. FNP uses a separate SRM to give re 3.in"d**,
o mad it operable within 30 oposed that they C hai would submit a report to the NRC if they lostgls"SRM days. The report would explain SNC's pre nned a at thod# ensuring the reactor remains in the shutdown condition in the t of a ntrol roheVacuation, the cause of the inoperability, and SNC plans and sc
'for restpnng the
% to operable status. The STS requires the plant to be in shutdow ition aftier 30 dayi since adding this SRM to the TS is more restrictive, and since this SR rovideshalind cd n only and is not used in any automatic actuation signal or t or the operati'onTiny component necessary to maintain the unit in ModId (hot stan
. nd SNCQifiEtion acceptable.
ITS 3.3.5 - L s o Power. Diese enerator Start Instrumentation SNC in their ITS c o,n modif ction 3.3.5 to include a new degraded grid alarm for the ' Loss Awe nstrumentation" section of the FNP ITS. SNC revised ITS Section 3.j3 and the cor asis sections to incorporate this change since the
- requiredget on and comp r this alarm is different from other functions in this section.
This alarm does not exist i TS. SNC revised the ITS because of a commitment made in resp % to a finding doc nted in the NRC Inspectic,n Report Numbers 50-348/92-17 and 50-364
. The NRC sta. accepted SNC's commitment in a SE dated November 21,1995.
Ba this, the st Y
s the change acceptable since it is more restrictive and meets the pre ent.
ITS 3.
Pressure. Temperature. and Flow DNB Limits ITS SR 3.4.1.4 contains a note stating that the 18-month RCS minimum flow surveillance is not required to be performed until 7 days after a 90% RTP. This note is consistent with Westinghouse STS SR 3.4.1.4 except that the elapse time is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> instead of 7 days in the STS. The intent of the note is to require performing the 18-month surveillance at the beginning of each fuel cycle with the reactor power as close to stable 100% power as possible, especially when performing a precision calorimetric heat balance measurement. FNP's CTS do not specify a time limit for performing the 18-month surveillance. With the added note, SNC will complete
26-q the surveillance within 7 days once the unit reaches 90% RTP, in accordance wit bple 1.4-3 in the STS Section 1, "Use and Application." SNC indicated that the 7-day limit is' based on operating experience.- SNC stated that 7 days pro e eno set up for measurement, allow for typical instrumentation problems, an
~ teve s ondiions without adversely affecting safety. The staff finds the 7-day limit to accepta Sii ITS.
ITS 3.4.2 - RCS Minimum Temperature for Criticality -
---l s Ted.6 c,mpWE 4.WmM Q FNP ITS S stziingTn'dTTitr1liTrveir ve,hyf ing "RCS Tavg in each
' loop 2 541' F" is required only if the Lo-Lo Tavg alarm is
'sii'd'any RCS loop Tavg is
< 547 degrees F. Using the Lo-Lo Tavg alarm in this not a p req is inconsistent with the Tavg - Tref Deviation alarm specigd in the for the surveillance i
ut is consistent l
with the STS in which either the Lo-Lo Tavg alarm d Tako - Tref alarm can be I
used. SNC provided the reasons for using the (pg"at median 7g awrm over avg - Tref Deviation alarm. The Tavg - Tref Deviation alarm is c
.urbin ading iWithis~ deviation alarm setooin ed u a and a default or low limited value for T f of 547 degrees F pri (if *5 degrees, ne loop Tavg could 1 des s F pri - #ctuating the Tavg - Tref
-Deviation aTarmcThe Lo-Lo Tavg al prpvides alarm a time the temperature in any of -
y the three loops is < 543 degrees F, th'eref s use is re conservative. In addition, i
three Trip Status Light Box indicati
~are avy to t, operator to indicate individual loop Tavg < 543 degrees F, eliminal neerns d9~efsihg, allure of the alarm. The staff concludes l
A 3.4 1.2. n yid M
- 52'-
W
' 2-sma s w.w
%pygg m p.4 g%.de' ITS 3A5WRCS Loops -
1 W
Re.5 tee s
-=
krequires,e_ ' ying steam g ne or secondary side water levels 4
are 2
(
. ran ops.
he CTS is 10% while the STS has a bracketed 17%.
)'g2, g
in respon a staff que
$ stated that the basis for the minimum steam ger;erator leve 3A.%2-during 3 is to ensure i steam generator tubes remain covered, thereby ensuring that t ssociated RCS loc pable of providing the heat sink for decay heat remov a
part e conversion to t p TS, FNP requested Westinghouse to det 5:" mesm
<> meet the basis stated for SRF3.4.5 3 estinghouse determi geni lev ecessa th ang m generator level was the bounding minimum leve The staff
! hat his lated boun ing lev is reasonably conservative an _ _ eindes that th rang $ steam generator leveQs acceptabD_ e s
Nc m'um level o 4 id in 74edI 5 EA V4 56As E*h) ~
new CTS 3/4.10 A - Reactor lant tooos g,4 g..g y9 CTS LCO 3/4.10.4 permits SNC to suspend CTS LCO 3.4.1.1 while performing start-up and
, physics tests when thermal power does not exceed the P-7 interlock setpoint (10% of rated thermal power). As a part of its ITS conversion, SNC proposed to delete CTS 3/4.10.4 from the CTS. SNC stated that this test exception was intended to allow reactor power operation up to 10% rated thermal power with no RCS loops in operation to circulate coolant. At FNP, this test exception served the single purpose of allowing SNC to accomplish natural circulation testing
during the initial plant start-up test program. There are no current requirergnts to n inue
~
natural circulation testing after the initial plant start-up test program, and EN_C d s not perform other testing at FNP requiring this test exception. The staff a rees withgN,C@ assessment and concludes that it is acceptable to delete CTS 3/4.10.4 from fCTS aWd should not appear in FNP's ITS.
ITS 3.4.15 - Reactor Coolant System Leakaoe Detection trumen ion in CTS 3/4.4.7.1, the RCS leakage detection systems that (e,r, red to be operable include a containment atmosphere particulate radioactivity monitoringifys either a containment air cooler condensate level monitoring system or a contaigment rggaseous radioactivity monitoring system. With only one of the uired I detection syst6ms operable in Modes 1 through 4, FNP operation m rupi
$rovided SNC obtains and analyzes containment atmospherey samjeg,least o every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the required gaseous or particulate radioactivginonitory ste is i rable. Otherwise, FNP must be in at least hot standby within next wurs shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
For proposed ITS 3.4.15, SNC revised the applL le STS
.15 LCO for leakage detection instrumentation to be consisten* with7NP'sS 3/4.4.
e STS LCO requires one e cofainment containment sump monitor, op'difai[ cooler cg 'i mo ere radioactivity monitor (gaseous or arfd 6ne containm nsIiitE flow rate monitor to be operable in particulate);8EI{tM$SNC wTt$[dMve 30 days tY estore an inoperable cont Modes 1 thr monitor. SNdw"o%dfave 30 dI%IE$ tore an inoperable containment atmosphere radioactivity mon'itbYkrTrify thatWeiEitIinment air cooler condensate flow rate monitor was operable. If the contNSEIll'atmosi ere"rYdioactivity monitor and the containment air cooler
$fi^
condensate flov(j$IeYoI1}iI re inopelable, then SNC would also have 30 days to restore either monit Since the CTS does not inc e requirement for a containment sump level or flow monitor, the oEEnal FNP SE Repoh[(NUREG-75/034, Section 5.6), the staff f det uipment a[ methods acceptable to satisfy the requirements of General Design Crityibn
'of 10 Cff Part 50, Appendix A without a containment sump monitor. FNP's contaT5niEilY.iiiiNesign prevents it from being qualified as 6 leak detection system per t
RG 1}5 ' ' '
Proposed ITS 3.4.15 would extend the time allowed to restore an inoperable RCS leakage detection instrument.to operable status to be consistent with the applicable STS. The CTS allows 7 days to restore the gaseous or particulate radioactive monitoring system, if inoperable.
The proposed ITS 3.4.15 change would allow 30 days to restore the containment atmosphere particulate radioactivity monitor, if inoperable, or 30 days to restore either the containment atmosphere gaseous radioactivity monitor or the containment air cooler condensate flow rate monitor, if inoperable. The associated actions for an inoperable required leakage detection
{
]
28-instrument (s) under proposed ITS 3.4.15 includes analyzing containment ap'e osp rab samples once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or performing an RCS water inventory bat nc very 24 ho' urs.. FNP would be required to be in Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, an de 5..
own)in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> if SNC did not meet the required actions and associat mpletion stated in the applicable STS.
-The staff has determineif that for one -inoperable leakage ion. nitor, suffi exist with one operable leakage detection monitor and th me'd' actions to anal containment atmosphere grab samples or perform an RC tory balance once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the inoperable monitor, to provide the necess to ensure that RCS leakage will not go undetected. Additionally, SNC stated t
point temperature monitoring system should be available to provide furthe suranceD _.
ill detect'RCS leakage in a timely manner. Therefore, the staff eq.
. Lthat extend fallowed outage time for one inoperable leakage detection monit et 30 da r proposed ITS 3.4.15, is acceptable.
-. ]
SNC added a new Action to the EC Sh'utdo
. CO. Tg ew Action provides an allowed outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the req,_
ECC triffu,c iarging subsystem to be inoperable
'provided the remaining operable ECCS com ne Wre' capable of providing 100% of the ECCS flow,
'uiQlent to a
[o" rable E ]f~or overpressurization ' concerns is notE operaDe ensure that ne centrif applicable unlil ratur re of the RCS cold legs is s 180 degrees F.
m Therefore, in M r more rifUgal charging pumps may be available and an Action
^
similar to that in ST S-O irfg') may be applied in FNP ITS 3.5.3 (ECCS-Shutdown).
The pro d allowed outa
.of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for this condition is consistent with the time curren allowed for one tr ECCS to be inoperable in Modes 1-3. The exposure of the unit to th small probability of vent requiring ECCS during this time is considered insignificant
-and 8fi' stet by the benefit ained through avoiding unnecessary plant transients.
1 This etion ovnies 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of outage time for a ECCS centrifugal charging subsystem whe*r(IE[ECCc@$
hutdown. However, this new Action is permissible only if the remaining operable.
omponents are capable of providing 100% of the ECCS flow, equivalent to a single operable ECCS subsystem. Another Action requires that if 100% ECCS flow equivalent to a single operab'.e ECCS train is not available, then the required action is for FNP to be in Mode 5 in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. In combination with the above stated requirements, the staff finds the proposed new Action for a 72-hour outage time for the ECCS centrifugal charging subsystem while in Mode 4 to be acceptable.
j
[
<4 29-ITS 3.5.5 - RCP Seal iniection Flow
'SNC proposed to change the way they confirm reactor coolant mp(
njection flow.
In converting to ITS, SNC would use a graph to measure se i o ection verifying a single operating point. SNC would determined appropriate based between the RCS pressure and the charging discharge he j r press The graph are based on FNP-specific safety analysis assumpt ns wh f vide the
~
between sealinjection flow, RCS pressure, and charging leader pressuretagra range of values for each of these parameters. Establishing we differential pressure allows SNC to more precisely and repeatedly verify sealin b d proper throttle valve position. The NRC approved this method of determiningM seal n
w limit for Vogtle.
In response to an RAl, SNC explained how they es grap ased the graph on a minimum differential pressure between the. argw:g
[andt essurizer and verified total seal injection flow to be within th
- its de E h.ac _
ance with the seal injection resistance assumed in the ECCS a ety an fes.
Q24:pm and 31 gpm points are based on the required flow and differenti Iressur termin
'JnVccordance with the.
conditions discussed above. The 27 -
oint i sed on linear graph between the previous points, which is conservati 6om to the al 27 gpm point that would be
- determined in accordance with the itions scussed ve. This information justifies the acceptability for using the grag ce of theTiEglep* $ rating point. Therefore the staff finds this proposed TS^ch_ange t be acce, table.
ITS 3.7.1 - Reducono the Pow r'RanoeWeutron Flux Setooint The ITS retains t S~
' men re, uce the power range neutron flux high trip setpoint based on pla nd the stinghouse Nuclear Safety Advisory Letter NSAL 94 recommen n
uary 20,1994. The trip setpoint is reduced to mitigate the loss of t ne load / turbine nt, since the high pressurizer pressure and over temper ure delta T trips m act quickly enough to prevent over pressurizing the secondary syst hen the initial trans ent condition is partial power operation with inoperable main steam sa Ives (MSSVs) uses a note to revise the applicability of this action in the ITS that ont s redu '
power range neutron flux high trip setpoint in Mode 1, and not in s th Mo TS currently requires. In Modes 2 and 3 a rod withdrawal event is the ont sie scussed in the FSAR, which would result in a significant power rise and pote nge MSSV relief capacity. The power range neutron low-flux trip and the source range neutron flux high trip provide adequate protection for this transient in Modes 2 and
- 3. Therefore, this change in applicability of the action to reduce the power range neutron flux high trip setpoint is acceptable.
ITS 3.7.2 - Main Steam isolation Valves C,d f
SNC's March 12,1998, letter proposed to revise
.7.1.
" Main Steam Isolation Valves (MSIVs)," to take credit for redundant MSIVs in each steam line at FNP, Units 1 and 2. CTS
W."7 1,6 Q4jJ nd STS 3.7.2 for MSIVs appi gl pres rized-water reactor esig ch include a single MSIV per steam line. In CTS gJ ith one MSIV ino b
ode 1, SNC has to restore it to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in Mod within With one MSIV inoperable in Modes 2 and 3, SNC has to restore the ope or close it within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entering Mode 2; otherwise, be in Mod
. the ne ode 4 within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
At the FNP, there are two MSIVs in each steam line, whic undant. One M
.ea steam line is designed to actuate on a train A ESF isolatio (
the other MSIVin each steam line is designed to actuate on a train B ESF isolatio g
o MSIVs are installed adjacent to each other on each steam line with no signifi pi
_ tween the valves.
The actuation of either MSIV in a steam line fulfills the sotation regatteriientsrgf the applicable safety analyses. There are no design basis accide s that re7 h MSIVs to close in order to mitigate an event at FNP. Since there A nce betwee _ e FNP MSIV design and the design the STS address, FNP chose. ake er SIVs in each s line for proposed ITS 3.7.2.
p p in proposed ITS 3.7.2, SNC includes tyf conditio (A and r inope ble N IVs in Mode 1.
- s have nopera MSl\\Mn Mode 1. The Condition A applies when one or mog ste'am li,Io ope a required action is to restore the inop"er;able MS status ithin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The 72-hour completion time is bastd on completi5nli'me provided in tt CTS and the STS for one probability o En$redundadkigered safe *t@ time that would requi inoperable train UIe'(ESF) syste ns. Also, there is a low Condition B(a$ lid %'sfo)'
accident occumng Buring this
'e the MSIVs to close.
r one MinErliTieam lines have two MSIVd inoperable in Mode 1. The required action iWoYe one hTSTVYEhrable Matus in the affected steam line within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The 4-hou IpE time iM Itent with the CTS and the intent of the STS requirements i$t$
n funcTon !n plants designed with only one MSIV per line. If
~
the requir ion and a mpletion time of Condition A or B are not met, then Conditi would require be in Mode 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Condition C is consistent w' CTS STS actions and the same power reductions and completion times.
y g in p ed ITS 3.7.2,-
, includes two conditions (D and E) fo ~
rable MSIVs in Mc des 2 an ndition D s when one or more steam line e a single inoperable MSIV(n Mo eguired action is to restore the i rable MSIV to operable status or close at the affected steam line wit '
days and verify that it is closed once every 7 day res
. SNC proposed this 7-day mpletion time since this would occur in Modes 2 and 3 when MSIV testing and maintenen including valve stroking, may be performed during hot conditions, and power is restricted to /. reactor thermal power. Condition E applies when one or more steam lines have two MSIVs noperable in Mode 2 or 3. The required action is to restore one MSIV to operable status or verify that one MSIV is closed in the affected steam line within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and once every 7 days thereafter. The 4-hour completion time is consistent with the CTS and the intent of the STS requirements for a loss of isolation function in plants designed with only one MSIV per line. If the required action and associated completion time of Condition D or E are not met, then Condition F requires the unit to be in Mode 3 in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 1
i i
I
Mode 4 in the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Condition F is consistent with the CTS a the S' lons and requires the same power reductions and completion times.
The staff concludes that SNC's p'roposal to take credit for re
'nt M S
.7.2 is acceptable based on the following:
The action times allowed by the CTS and the STS f
.singtp noperable consider that a redundant MSIV remains operable i Fiffd,ed steam line a capable of performing the intended safety function as
. N design
= There are no design basis accident analyses that w dr gre MSIVs to close in order to mitigate the event.
ITS 3.7.8 - Service Water System SNC's March 12,1998, letter proposed to a 'an ac Service Water System (SWS)," to account for the redundant aut
$ tic turb build tion valves in each service water train at the FNP, Units 1 and 2 CTS 3/4.7.4 and the associated STS do not account f pecific design feature at the FNP, which includes two redundant attordEtic turbinT6Gldi lation valves in series in each SWS train. The valv$'close autor$atiE5kon a sa 5tMnJEElion signal to isolate the non-safety turbine l
building SW$To"$$iFand ens 7rT5ifeguate SW$ flow to essential components. When two of these isolatioYv'aivTi(o(e in eYc(ditioEtha't would be applicab!a There Sgtrain, become inoperaole, the actions in both the CTS and the STS doi 8t 3ddress a con
~
LCO 3.0.3. SNC beliEB3f 0?at.enterEn^[LYD,3.0.3 in this condition would be overly conservative since there are twolu't'oNajIEiDrbine bMjEfing isolation valves in each SWS train, and one automatic vafv'e in each lIdidkould still remain fully operable. Therefore, two 100%
capaci NS trains would "Yvailable to provide the required system safety function.
in proposed ITS 3.7.8, S dded an action for one inoperabie automatic turbine building isolafi6%n~ alve in each S train, which requires that both inoperable valves be restored to oper56i'* status in 72 ho8Is. SNC based the 72-hour completion time on the fact that the e
isolaiiEEfuncjign,.p'erfo5ned by the'se valves is not lost in either SWS train since one automatic turb%yh(ng,o1"aTion valve in each train remains operable. Although the reliability of the isolatiohJunction performed by the automatic turbine building isolation valves is reduced, there are still two 100% capacity SWS trains available to perform the required safety function of the system, and there is a low likelihood of an event occurring during this time that would require the isolation function provided by these valves.
Based on the above, the staff concludes that adding the action to ITS 3.7.8 at FNP is acceptable based on the redundant design of the automatic turbine building isolation function in each SWS train and to prevent an unwarranted entry into LCO 3.0.3 resulting in an unnecessary plant shutdown.
-32 A
ITS 3.8.1 - Removal of Accelerated Testina Reauiremen'ts for Emeraency. Diesel Generators
~ CTS surveillance 4.8.1.1.2.a refers to accelerated testing req ments gemegency diesel
. generators (EDGs). Table 4.8-1 of the CTS contain these a ted test
' ments.
Table 3.8.1-1 in NUREG - 1431 (STS) also contains EDG rated te nts.
For the FNP ITS, SNC proposes to delete the EDG accel test requirem with GL 94-01, " Removal of Accelerated Testing and Spe
- ng Requireme Emergency Diesel Generators" guidance. GL 94-01 contai nce that allows utilities to remove the accelerated testing requirements specified in provisions of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of en assure EDG performances are implemented. After
- piem,d,ance w,_
4 clear Power Plants (maintenance rule), including the applicable regulatory gi Mgide a program to e ting the 'Ije formance program as required in the mainterience rule, G -
.no es hat utilit' Ban remove individual-EDG accelerated testing requirements from
- ince
.as
. nted the maintenance
. rule, the proposed ITS do not include acce
- e.d tes '
req
. However, RG 1.160,
" Monitoring the Effectiveness of Maintena at Nu ar Po s," addresses a docketed
' ommitment to maintain selected tar '
, bilit" a ues for fGs. In this regard,'the staff
~
c expressed concern regarding such k'eted mitme n response to this concem, SNC noted their commitment for an EDG, rget re!
'r va
.95. Thus, by implementing maintenance rule requirementsfSNC' effectiv
.rels1 he intent of the accelerated testing requirements:tE5nsure EDG'rkliitiility and av $S0tfand as such the regulations continue to provide ade46dWEiIisurance-D5" ormaEThe frequency of ITS EDG testing is 31
~
days as deterQMfrYm STS
. This frequency is also consistent with GL 94-01 guidance. Based drEtMibove, t ja acceptable.
.8.2.1 - E 1 red Surveillances for Modes 5 and 6 Plant Conditions That Demonstrate 5Cacabilities'NdtiR'salired for These Modes
~ that deY.8onstr' ate capabilitieYthat are not required for these Modes. Table 1 sho 4
2 requires'SNC t rm certain surveillances in Mode 5 and Mode 6 (refueling) surye[i[l nce requiremengt f(ERs) and the corresponding ITS SRs. CTS SR 4.8.1.1.1.b addr,essktransferri me unit power supply from the normal power circuit to.the alternate cire 8 addresses a simulated safety injection signal overriding the EDG test mo addresses simultaneously fast starting of each EDGs. SR 4.8.1.1.2.c.3 addre o natic fast starting of an EDG from a safety injection test signal (without loss of effsite power). SR 4.8.1.1.2.c.9 addresses EDG automatic load sequencer timers. SR 4.8.1.1.2.c.4 addresses simulating a loss of offsite power in conjunction with a safety injection test signal and automatic EDG start and load sequencing.
. ~
Table 1 Surveillance Requirement.
CTS RS
--- - SR 4.8.1.1.1.b -
hR 3.8[
SR 4.8.1.1.2.c.8 kd 15 SR 4.8.1.1.2.d
[h SR 4.8.1.1.2.c.3 SR3.
SR 4.8.1.1.2.c.9 M 3.8.1.1 T SR 4.8.1.1.2.c.4
[
M,1A[
t
~
.2 su ancest iminate t requirement to perform the above ITS Section 3.8.1 surveillances for 5 an plant c itions. Since the SRs for AC sources - shutdown define and r
he oper yJeg rements of the AC sources required in tdown s05e"'iilarIEe requirement is revised in the ITS to Modes 5 and he AC source s
the app @EeTr3h,grability red 0frEnents. The definitio
,i T~
more clearl g
ent b pabfe of performing its required safety function. The above to the system surveillances pro n the ITS-
, tions for Modes 5 and 6 do not demonstrate any capability related requiled safet
_n,,of an AC source for these Modes. The revised AC sources - shutdo_wddiGiliEces in t do not require SNC to meet,or perform the excepted surveillancesidr Wlod t conditions. SNC took exception to surveillances that require t owing:
(1) emonstrating the p ility to transfer offsite circuits (only one offsite circuit is required
{
n Modes _5_ and
(
l emonstratin source response to an engineered safety features actuation (SI)
I($1i a required safety function in Modes 5 and 6)
(
starting independence (only one EDG is required in Modes 5 and 6) cept
.8.1.10, eliminating the requirement to perform the above FNP ITS survelliances for Modes 5 and 6 plant conditions is consistent with NUREG -1431 guidance. SR 3.8.1.10 i
verifies that an SI signal fast starts each EDG, each EDG operates for.y 5 minutes, and emergency loads are energized from the offsite power system. However, for Modes 5 and 6 plant conditions, operators defeat the SI signal. In this regard, requiring SNC to perform this surveillance for Modes 5 and 6 plant conditions demonstrates a capability that is not required for these Modes.
(4.) vefihij th,aub M c. b>A M P'**'t % i cp.M d emaynq WA seymtm i
, MM W in addition, the FNP ITS Bases ole that during plant shutdown Modes, a consiste with ITS Section 3.8.10 (Distribution Syi rrsr-Shutdown), portions of a second tr th(electrical power distribution subsystems lequired to be operable. Further, th NBases note the following
' ' dire %e, gM 7h J,
7 J *.5ec
- 4vo h 40 be. OPGtA G (1). Required portior)s of the seconMrrot aTerna current subsystems may be energized from the associat vert onnected t u
direct current (DC) bus or the attemate Class 1E rce consisting ol4Ein0ert r static transfer switch and the associated constant former, or g3g-(2)
Required DC buses associated with the second
, subsystems are energized from either an operable DC source sting o"Ytg35 tery, one baftery charger, and the corresponding control e interc iig cabling associated with that train or a battery a,rge rres ing control equipment and the interconnecting ng wit The staff expressed concern regarding liabilit a less liy complimented second Class 1E power train, since the secon ertr may notynsist of a fully complimented Class 1E power train. In response irconce SNC a reed to include in the FNP ITS Bases Sections 3.8.8 (Inverters - Shutdo and 3.8.
he g-MM4Absle_ms-_5hdebm A
Class 1 r and distri stems are nyally usedYecause these systemsTre to eve'Nts such as maintenance or modification,
.availabt
_ ii'able. H*
~
portions o (d[i,ay E sysYegggemporarily unavailable. In such an instance the plant staff assesse emate systems to ensure that defense in depth is maintained and that risk The FNP o not req
, assess necessary second power trains during plant shutdowg odes to minimi s such, these actions are considered voluntary. In addition, '
licenseG' voluntary actions the CTS, which include safety planning and assessment in shut
, were an impo part of the Commission's decision to cancel the shutdown rule.
Ba the above, onclude that eliminating the requirement to perform the identified su orM 5 and 6 plant conditions is consistent with thyrovided in NUREG -
143 jd is acceptable. Further, we conclude that SNC's iTe untary actions for the necestery.secon power train during plant shutdown Modes are consistent with the Commission's decision to cancel the shutdown rule and are acceptable.
ITS SR 3.8.2.1 - Addina a Note to Exclude Performina Specific Aeolicable Surveillances On Certain AC Sources for Plant Shutdown Conditions ITS SR 3.8.2.1 adds a note to revise CTS SR 4.8.1.2 for AC Sources During Shutdown. The note identifies ITS SR 3.8.1.8JR 3.8.1.11, SR 3.8.1.12, SR 3.8.1.13, SR 3.8.1.14, and SR 3.8.1.18 as applicable but not l equired to be performed. SR 3.8.1.8 addresses each EDG
6R 1 bed Me bh
"#b'"1
- h
'.,t '.
af %t EDG
- o. \\oss of oW
- par, y
-3 5 -
rejecting a load greater than or equal to its associat'ed single largest post beiden R
3.8.1.11 addresses bypassing ED_G protective trips when receiving n
$11
. SR 3.8.1.12 addresses the EDG 24-hour endurance test. SR 3.8.
~
ch EDG k 1_0 minutes,
starting and achieving specified voltage and frequency valu 1 se of shutting it down after operating it for2 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> loaded a reat the rg dts1Kpecmed load. SR 3.8.1.14 addresses synchronizirig a loaded EDG, th th ogia po transferring the EDG loads to the offsite power source. S ). 1.
Tddresses e re'e i a los_d of21200 kW and5 2400 kW.
o A johJ fws en s **
Adding thiinote as : e e o op req in e opera e to..
alleled with the offsite power network or otherwise rendered inoperable to perfospi a su 1_
ASome of the CTS surveillances required to be performed involve tests tha uire paheleirinhetEDG to the offsite power network. This condition, with the on DG andy gpy required offsite circuit connected, presents a risk of a single fau station bl Rout. To address this
- concem and to avoid other conflicts with testi an'd o b Mhje neludes a note with SR 3.8.2.1 to exclude the requirement to perfor certain rve !
Etestr. The exception provided by the note does not exclude the require nt for t
'particular function but rather SNC does$ ' have G to da able of performing the emonstra e the capability while the EDG source of power is being relied on toget'the
- d. Thusj, [dding the note to the FNP ITS SR 3.8.2.1 to exclude the requirement rform above4urveillances is consistent with STS guidance and i eptable.
ITS SR 3.8 Alldina a Decific Surveillances Reauired to be Performed for Modes 5 FNP ITS SR 3.8 pe notesthai R 3.8.1.1, SR 3.8.1.2, SR 3.8.1.4, SR 3.8.1.5, and rformnem for Mode 5 and 6 plant conditions. SR 3.8.1.1 SR 3.8.1.6 a an ~
verifies c reaker a n indicated power availability for offsite power circuits.
verifies that eac ris from standby conditions and achieves specified steady-state ge and fre_quene aYues. SR 3.8.1.4 verifies that each EDG day tank contains a minimum specific number allons of fuel oil. SR 3.8.1.5 verifies that the fuel oil transfer syst "perates to transf uel oil from'the storage tank to the day tank. SR 3.8.1.6 verifies tha EDG start a standby condition and achieves specified voltage and frequency val The CTS.
1.2 provides requirements for the AC Sources during shutdown by referencing SRs 4.8.1.1.1 and 4.8.1.1.2. The CTS SR 4.8.1.2 only provides a specific exception to the referenced surveillances required for the shutdown AC sources but does not identify the specific surveillances required to be performed. The STS provides additional exceptions, as identified and addressed above, but also does not identify the specific surveillances that are required to be performed for the shutdown AC sources. Proposed ITS SR 3.8.2.1 adds a specific list of surveillances that apply and which must be performed. This list of required st peillances is consistent with the FNP ITS lists of surveillances remaining afterceleatinoixclusions which are not required to be performed. The added list of five required surveillUices for ITS SR M ev'
F, 36-3.8.2.1 is also consistent with the five surveillances the STS require to be rforme
- hus, adding this list of five required surveillances does not introduce a techni the SRs, but'rather is an administrative change which is provided to clea identijf eillances actually required to be performed an the shutdown AC sour
'ecordi nge is only an administrative change that{
- he five specific surveilla s require ed and thus is acceptable.
ITS 3.8.4 - Revisions to STS 3.8.4 Actions to Retain An k N Footnote and erv Connection Resistance Reauirements kdesign and CTS ITS 3.8.4 actions revise the STS 3.8.4 actions to be cons t
requirements. ITS 3.8.4 provides new action conditioni ed on G
d the FNP specife rol poweg 5osed FNP design which includes batteries dedicated to the S s
specific action conditions are necessary to retal FJ c allo provided by an asterisk (*) footnote associated with two of th TS s
~
ddress the fact that the inoperability of the service water intake cture the associated train of the SWS.
An asterisk (*) footnote to the survei of bM the aux building and service water intake structure battery CTS provid an allowano's to fdeclaring the battery inoperable due c
to connection rysistance not wit 6in it. Thi blishes a 24-hour completion time to restore connection resistanceT[EI n the requi rance. The CTS allowance to restore battery conde'EtiI istan ir'ghe requir*e'~d tolerance is supported by lEEE-450 which notes that con stan
.me n indication of conditions that can be easily corrected prior t pect ition, IEEE-450 does not note that battery connection resist. _.
ecd is on eclare the battery inoperable. In this regard, ITS for connection resistance not wittiin limits. The 3.8.4 includejs SepYra o -conditi completion time associate pnt? Lproposed ITS condition is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> which is the same as that spe in the CTS f ITS action conditions vise the STS 3.8.4 action conditions to incorporate a specife def ion for the se water intake structure battery system. This change is necessary sin Ebattery syst upplies only one required safety system, which is the SWS. The S
a:co time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for an inoperable train. Each train of service water inta les supplies the associated service water DC control power. Considering that e th units at the same time are permitted to have one train of service water inoperable for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (as permitted by the CTS or STS) fo-wns other than the DC control power, the completion time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with a plant shumowri.
red if the completion time is not met is very restrictive for an inoperable service water intake ss.ucture battery train. In this regard, ITS 3.8.4 provides separate action conditions for an inoperable service water intake structure battery' train which, if not met, default to a condition that requires the associated train of service water to be declared inoperable. This action condition format is consistent with that contained in the STS where similar support / supported system relationships exist.
s )
ITS 3.8.4 also revises STS 3.8.4 battery connection resistance SRs to be consistent h the FNP CTS.. The terminology used in the surveillance requirement and the$i,s values specified are revised to be consistent with the language and corinection Ksistiiince the FNP CTS. The above changes are consistent with the S orf' Efe' sign, are provided in a format and presentation consistent with that ained in t
!a acceptable.
e ITS 3.8,9 - Revision to CTS Action Statement for Service rDC Distribution an ery Systems CTS 3/4.8.2.5 action statement for the LCO for the servictwate ion system requires that with one of the 125-Vdc distribution trains inoperab,legestore no rable distribution system to operable status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in a.l(astTiot standby e next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Pr YifST8g. revise CTS LCO action statement such that with one of the 125-volt dift'r butiorftfaTrIdinopera
,[ restore Ine inoperable i
[ distr Dution system to operable status within'2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> arid 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />,from discovery of failure to (meet the LCO or declarefthe associated sirvice walditrai in" r ble_immedig ely.
)
At the FNP, a separate DC distribution and batte,[ry system,at is Y
ts Stc.IoA*A distribution and battery system supfies SWS%, control, wer. The primary purpose of the service water igtake structure DC diMribution in36attIry system is to supply DC control power to g
the associated service waterIrMihe servicEEa't5Mtake structure DC distritution and battery system doe 6IdQply any*otgrgrelated olYequired loads. The CTS requirements that apply to an inoperakservice water train aiIo$rJ2 hours to restore that train to ope ~rable status before requiring a plant siftftEIdi As thNESIEEwater intake structure DC distribution and battery S olely toIE poit the associated SWS train, the CTS allowance of only 2
{
systems are require'cIh{gpr intak[e1 structure DC distribution and gEEerv hours to resto status befor requiring a plant shYdown is very conservative. If an entire service water train may be inop a e for any reas(f 67o 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, it is very restrictive to require a plant shutdown to begin hours if the DC cgntr power to the same service water train is inoperable. The 2-hour rest son time associated with distribution and battery systems.is based on the fact that DC sup Jgdistributiog[take structure DC distribution an syst ypically suppo any TS required engineered safety feature systems and the loss of the battery system impacts many required systems. This is not the case for the servicegagte.in actiog ements~ revise the CTS actions to be similar to other STS actions for support systems when tnaTSIIp*pYrt system becomes inoperable (that is, declare the supported system inoperable).
The current default actions to be in Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 5 in the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> are replaced with an action to declare the associated service water train inoperatile immediately. In this regard, the completion time for the support system becomes more consistent with the corfipletion time for the supported systemJhe c i 6 action to restore the inoperable distribution or
]
battery system to operapie status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is retained to ensure prompt attention to problems
'with the distribution or battery systems and also allow a reasonable time to restore minor problems before requiring that the associated train of service water be declared inoperable. In addition, Lretention of the CTS 2-hour action statement maintains creater consistency with associated STS
4 ~
factions for other distribution and battery svstame in STS LCOs On the basis of the ove technical reasons being consistent with the guidance provided in the STS
" sed action statement for ITS 3.8.9 is acceptable.
ITS 5.1.2 - Administrative Controls - Resoonsibility SNC proposes to adopt less restrictive wording for a portion of TS Responsi or ing contained in their submittal would substitute the generic titEf, Senior Reactor Oper
' SRO) for the ITS specific title of Shift Supervisor (SS). The specificati6n"rd$ ires that an individual must have an SRO license to assume the specific command and[ofidotDction. An on shift SRO can be designated to assume the command and control functi$rIwh7Ei 5$1 eaves the control room because the qualifications, as defined in ANSI 3.1,1993 " Selection Personnel for Nuclear Power Plants," for an on-shg*O and for the
@desgriated as the Shift Supervisor are the same. The change in wording _is fes res ' ive but c"oYs'not reduce the underlying requirement of the specification a dMerefo ifcle abid[
ITS 5 3.1 - Administrative Controls - Unit Staff Oualification d
SNC proposes to adopt alternative wo#3 rdin'g to refer to the
- idual responsible for management of the radiological protection program [o[ plant o[lili$tionsg'.
proposes to use the generic the RadiatiopfI8tection Man'a'g,r@phe geneNEtit reference of " senior individuali harge of Hea les in lieu of the ITS specific reference to e
would apply to 5.3, Staff Qualifications. The use of a generlEf5firence lIflieu,r.ofYspecific iI oes not change SNC's current commitment to an appropriaEe%$lifiE5 tion sta71kdgspecified in TS 5.3. Although the wording proposed by SNC does not mdIEiiiEITS specific reference, the individual with responsibility for the radiation therefore, is acc[eptd9eaYbEIdentifiedY 1 Yequired to be qualified to an a protection progra ITS 5.5.7, 'RCP Flywheelinspection Frecuency (We wiGrovide the SE inp[at a later date[phitel SE ted a
Wy AV
~
ITS. 6' >- Revisions to <.he Emeraencv Diesel Generator Failure Report
.Y W
The annualNel generator reliability data report in CTS 6.9.1.12 is replaced in the ITS with STS 5.6.7,TD575IiuTe Report. The content of the replacement report is modified to include additional information to be supplied with each report consistent with the current FNP practice. The CTS annual EDG report requires that all tests and the number of failures to start on demand for each EDG be submitted each year. In addition, the CTS requirement references RG 1.108 for report content. The STS EDG reporting requirement is based on the number of failures in the last 25 demands, and a report is only required to be submitted when an individual EDG experiences four or more valid failures in the last 25 demands. The EDG reporting requirement is revised in the ITS to correct the reference for the additional information to be included in the report. The additional information to be provided in the EDG failure report is currently derived from the FNP EDG
Eo..
reliability monitoring program, which provides more information than the CQ refe to RG 1.108, This FNP program is referenced in place of the STS referencQTo R
.9 and 1.108 for the additional information and exists to fulfill a commitment idediqg a to the station blackout rule. The elements of the EDG monitoring progra
, BonsisI. f t guidance provided in RG 1.155. RG 1.9, and Appendix D of NUMAR 00,Rev
?
~
onitori
~
program ensures the data on all EDG demands is logged evalua ndt ia performance is monitored in accordance with RG 1.155, F
,a, ppendix NUMARC 87-00, Rev.1. In addition, this monitoring progi
~. ifes the actions '
en if one or more of the EDG reliability performance indicators
' e program reaches or exceeds the FNP reliability trigger values. In this regard, lure report is revised in the ITS to be consistent with the current FNP practice re malinformation to be included in the report. The other aspects of the STS equire more relaxed -
requirements compared to the CTS requirement be su ch year regardless of the number of failures and all tests and all f
.I d eac ar. However, the revised reporting requirement does require ali s a description of the failures, underlying causes, and corrective n tak information is necessary to s.,
asses _s_the EDG reliability and the overall ctiver ofth NP' DG maintenance and testing program. Therefore, the STS reporti ire Ias mod in the ITS by reference to the FNP EDG reliability monitoring program i (ptab it con Es to provide a means to rnonitor the FNP EDG reliability and allows for ctive ures taken if required.
ITS 5.7.1.c ninistrative ois - Hioh lati Area SNC propose oa o alterna e woy in TS 5.7.1.c, High Radiation Area. The ITS references
~
The existing FNP speElt3ca Bn.gerT, the Radiation PrMon klana g the frequency of periodic radiation surveillances.
assigns esponsibility to the individual with the title of Health Physics Supe #Fiif rsionTIITS, SNC proposes to substitute the more generic reference of ealth physi r rather than referencing a specific title. The~ requirements related t
' igh radiation al sin unchanged. The level of the individual required to establish trie fr
[ncy of radiation s nces also remains unchanged. SNC's proposed wording varies the re"uf:lTS specific refer me but is consistent with the existing specification and does not change from ements related a high radiation area and, therefore, is acceptable.
k IV. -
A E CONS ATION in acc shee with the Commission's regulations, the Alabama State official was notifed of the proposed issuance of the amendment. The State official for the State of Alabama had no comments.
V. ENVIRONMENTAL CONSIDERATION Pursuant to 10 CFR 51.21,51.32, and 51.35, an environmental assess ent and findi of no significant impact was published in the FEDERAL REGISTER on@ (64 FR Accordingly, based upon the environmental assessment, the Commiss on has d ermined that Insd ppv d.h A nuda
s*
Issuance of this amendment will not have a significant effect on the quality the h an environment.
VI. C O N C L U SI O N
!The FNP.lTS provide clearer, more readily understandable requirern 1 oe n
' of the plant. The NRC staff concludes that they satisfy th guidance n the Comm y
statement with regard to the content of TS, and conform t Te" ' el provided in N
~ f431
~
with appropriate modifications for plant-specific considerati
.I NRC staff further concludes that the FNP ITS satisfy Section 182a of the Atomic Energ
' 50.36 and other applicable standards. On this basis, the NRC staff concludes that t prop S are acceptable.
The NRC staff has also reviewed the plant-specific ACTS d in this evaluation.
On the basis of the evaluations described herei
.ei hang NRC staff concludes that these changes are acceptable.
The Commission has concluded, based le co ' ration ssed above, thati (1) there is reasonable assurance that the health afety e publi ill not be endangered by operation in the proposed manner; (2) such a 6 will
-. conduct compliance with the Commission's regulations; and, (3) the issuance e amen ts wili be inimical to the common defense and security or to the health
- s y of the i
Principal C uorsiC.SI hu H. Balukjian
. Wes az M. Padovan ader G. Hsil an rg.
C. Liang na M hiey Date: A t6,1999 4
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e em b
g e
CTS Discussion of Change Tables 9
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