ML20024G752

Forwards Request for Change of Tech Specs,Including Proposed Changes to Reporting Requirements Section Per Dj Skovholt to AV Dienhart

ML20024G752
Person / Time
Site:	Monticello
Issue date:	11/20/1972
From:	Mayer L NORTHERN STATES POWER CO.
To:	Giambusso A US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 9104290330
Download: ML20024G752 (1)

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4,e NORTHERN 5TATES POWER COMPANY

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Mr. A Giambucco k

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United Statec Atomic Energy Commicc Lon C"

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Dear lir. Giambucco:

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MONTICELLO NUCLCAR GENERATING PIMIT DOC}22 NO. 50-263 LICE' ICE NO. DPR-2._

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CHAUGE liEQUEST DATED HJ/ DOER 70, lo72 Attached are three cigned originals and 37 confomed copiec 01' a request 7 "

for a change o: Technical Cpecificationc c: Appendix A of the Monticella Nuclear Generatirr Flant.

Thic change request has teen reviewed ty the Menticello Oneratienc Ccamittee and Cafety Audn

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A regular cubmittal e r thic type normally requirec only three cig:.ed PD originalc and 19 conforned copiec per 10CFR50 59(d).

Thic requert for 97 change incluaec, in part, the propoced reviciens to the reporting; re-

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quirements section of the Technical Cpecificationc requccted in Mr.

u Donald J Skovholt's letter of September 23, 1972 to Mr. A V Dieahnrt.

.5 The 37 r'opiec attached are reopencive to the inctructic,c containea in

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GUITED STATES ATOMIC ENERGY COSMISSION NORTHERN STATES POWER COMPANY m aticello Nuclear Generating Plant Docket No. 50-263 REQUEST FOR AUTHORIZATION OF A CHANGE IN TECHNICAL SPECIFICATI0PS ct OF APPENDIX PROVISIONAL OTEFATING LICENSE NO. DPR.22 b

( Change Pequest Dated November 20,1972)

Northern States Power Company, a Minnesota corporation, requests authori-zation for changes to the Technical Specifications as shown on the attachments labeled Exhibit A, Exhibit B, and Ex.hibit C.

Exhibit A describes the proposed o.

changes alorg with reasons for change.

Exhibit B consists of copies of pages C

of the '5chnical Specifications marked up to indicate a portion of the proposed charges.

Exhibit C consists of newly preparea pages of the Technical Specifi-pc cations which present the rest of the proposed changes.

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'Ihis request contains no restricted or other defense information.

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NORTHERN STATES P C01 ANY *

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By o

E Arthur V Dienharf 90 Vice President-Engineering (if On this M S day of me2/d,1972 before me a notary public in b'

and for said County, personally appeared Arthur V. Dienhart, Vice President -

Ma Engineering, and being first duly sworn acknowledged that he is authoriced to execute thic document in behalf of Northern States Power Company, that he has c

c-read it and knows the contente thereof, that to the best of his knowledge, I

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information and belief, the statemente made in it are true and that it is not hf-ir trposed #cr de ay.

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g ohn J(fmith 43 Notary Public, Hennepin County, MLnnesota y*

JOHN J. SMITH gA. % % Henne,,n Caecy, Enesee frF

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}buticello Nuclear Generating Plant Docket No. 50-263 c:e s

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REQUEST FOR AUTHORLATION OF 4.R A CHANGE IN TECHNICAL SPECIFICATIONS OF APPENDIX A PROVISIONAL OPEFATING LICENSE NO. DPR.22 g,

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( Change Request Dated November 20, 1972) b n

Northern States Power Co=pany, a Minnesota corporation, requests authori-N-

zation for changes to the Technical Specifications ac shown on the attachmentc b'

labeled Exhibit A, Exhibit B, and Exhibit C.

Exhibit A describes the proposed i

changes along with reasons for change.

Exhibit B concists of copies of pagen of the Technical Specifications marked up to indicate a portion of the proposed 3 p chnnges.

Exhibit C consists of newly prepared pages of the Technical Specifi-4 cations which precent the ract of the proposed changes.

This request containc no rectricted or other defence information.

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NORTHERN STATES POWER COMPANY By /s/ Arthur V Dienhart Arthur V Dienhart g{.n Vice President-Engineering eo

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On thic 20 day of November 1972 before ne a notary public in

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and for caid County, perconally appeared Arthur V. Dienhart, Vice Precident 9,

Edneering, and being first duly cworn acknowledged that he 10 authorized to d,

execute thic document in behalf of Northern States Power Company, that he bas 9

mad it and knows the contents thereof, that to the best of his knowledge,

1. 7 inforT:ation and belief, the statemente made in it are true and that it is not interposed for delay.

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/s/JohnJSmith John J Smith w

Notary Public, Hennepin Coanty, Minnesota NOTARY SEAL Itf Commicsion Expirec thrch 3,197t m,

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EXHIBIT A

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E MONTICELLO NOCIIAR GENEFATING PLANT M

DOCKET NO. 50-263 CHANGE REQUEST DATED NOVEMBER 20, 1972 PROPOSED CHANGES 70 THE TECHNICAL SPECIFICATIONS P

APPENDIX A 0F PROVISIONAL OPERATING

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LICENSE NO. DPR-22 g!h

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- Q' 4 Pursuant to 10 CFR 50 59, the holders of the above-mentioned license hereby N.

propose the following changes to Appendix A, Technical Specifications:

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Proposed Chance b

On page 81, Section 3 3.E, change ".... Specification 6.6." to read Q

".... Specification 6.7. B. "

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On page 90, Section 3.4.B.2, change ".... Specification 6.6.E,... "

, g to read ".... Specification 6.7. C.2,...."

On page 131, Bases 3.6 and 4.6, change ".... Specification 6.6.E.3.d."

to read ".... Specification 6.7. C. l.d. "

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h On page 136, Bases 3.6 and h.6, change ".... Specification 6.6.E.3. "

to read ".... Specification 6.7.C.l.c."

Reason for Chance

' Substantial revisions to Section 6.0 of the Technical Specifications are pro-s

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posed later in this change request. he above revisions are necessary so that the references confom to the revised paragraph numbering in Section 6.0.

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Proposed Cbance

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On page 106, Section 3 5.H, delete the whole paragraph "When it is determined.... out-of-service period. " and replace it with the fol-

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"When it is determined that =aintenance to restore components 4-or systems to an operable condition vill last longer than the periods specified, the reporting requirements of 6.7.C.2 shall be fulfilled."

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Beason for Chance 9 (I Current paragraphs 3 5.H and 6.6.E appear to require two separate reports on the same subject (i.e., extended maintenance of certain components and systems ).

(f "he determination" as referenced in each of these paragraphs probably vill not o-be made until some portion of the allovable out-of-service period has elapsed.

' p" he due date for one report is prior to the end of the allowable out-of-service Period (seven days in most cases) and the other is due within se"en days of

.9 "the detemination. " 7his inconsistency or apparent dual reporting requirement 36:

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DGIBIT A (Continued)

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plus a desire to collect all reporting requirements and their definitions in Section 6.0 is the reason for mquesting this change.

The proposed wording for this reporting requirement is contained in Section 6.7.C.2 of this change request.

Proposed Change Page 1, Sections 1. 0. A,1. 0.B, cnd 1.0.C, is to be replaced by pages 1 and 1A, Sections 1. 0. A, 1. 0. B, and 1. 0. C, as contained in Exhibit C.

Pages 192 - 211, Section 6.0, Administrative Controls, are to be replaced by Pages 192 - 220 as contained in Exhibit C.

Beason for Chanze

'Ibe Technical Specifications, Appendix A of Provisional Operating Licenst No.

LPR-22, have been in effect for over two years.

Certain portions of Section 6.0, Administrative Controls, have been difficult to interpret, particularly Table 6.1.1 which tends to put the review committees in a line function rather than in a review and audit function.

Our operating and reporting experience under Section 6.0 of the 'Ibchnical Specifications has also identified portions which could be refined for clearer interp atation and improved workability.

Certain changes in the structurcs of the organizations, shown in Figure 6.6.1 and 6.6.2, were implecented in 1972 as reported in the NSP letter of thrch 10, 1972 to Dr. Peter A Morris.

During the past year, Directorate of Licensing personnel have expressed a desire for certain changes to thie. section of the Technical Specifications. A letter dated September 28, 1972.? rom Mr. D J g

Skovholt of AEC-DL to Mr. A V Dienhart of NSP requested that vt. submit, within sixty days, proposed revisions to the reporting requirements in the Adminis-CD trative Controls of our Technical Specifications which meet the guidance set iis.

forth in the enclosures to the subject letter.

NSP representatives have participated in a number of discussions v1 Lh AEC-Licensing and AEC-Regulatory Operations personnel in recsrd to a more acceptable and somewhat " standardized" format for this section of the Technical Specifica-

%p tions.

Based on these discussions and on in-depth guidance from the ANS Pro-k posed i tandard for Administrative Controls for Nuclear Plants, Section 1.0, Ibfinition of Abnormal Occurrences, and Section 6.0, Administrative Controls, N

have been rewritten n's contained in Exhibit C attached.

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'Ihe September 28, 1972 letter from Mr D J Skovnolt on reporting requirements requested that additional inform? tion be submitted on those Technical Specifica-g tion revisions and facility measurements for which it is not practical to follev the guidance set forth in that letter. We believe that Section 6.7, Plant W

Reporting Requirements, in this change request meets the intent of the guidance g.

provided.

(M The guidance given for Section 6.7. A.2.g on environmental monitoring reporting in the semi-annua'. operating report is more extensive than that required in Q

our current Technical Specification.

Environmental monitoring sampling can ry continue right up to the end of the six month period. Sample analysis, tabu-lation and statistical analysis of the data, evaluation of the data, and prep-aration of a final draf t of the environmental monitoring section of the semi-a J4 annual report in the detail proposed within the 60 day time period allotted D

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i EXHIBIT A (Continued) t p.,v 7

appear difficult, if not impracticable.

We believe that the annual Environ-mental Monitoring Report is the more appropriate location for this type of y"

statistical analysis and evaluation of its significance, particularly since C

it treats a whole year with a full cycle of seasonal variations as opposed to fg,.

the environmentally non-related six month operating report perioca.

If the semiannual operating report were to report environmental monitoring data only, without detailed analysis, this could be practicably accomplished in J.

the 60 day time period.

In spite of this apparent difficulty, we have used h('

I the guidance provided by the September 28, 1972 letter for Section 6.7. A.E.c of this change request and vill evaluate the extent of the t,'Arden imposed

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following the first applienble reporting period.

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vided that:

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Die component is returned to an operable condition within 7 days, or r (o.l.C.R J

2.

A written repo na specified in Specification Q chall be sub-mitted to the Atomic Energy Commis-sien when the mintenance to restore the component to an operable condi-tion vill last lonrer than 7 days.

C.

Volume-Concentration Requirements C.

The availnbility of the propar boron bearing solution shall be verified by Die liquid poison tank shall contain parformnce of the folloviru; tests:

a boron bearing solution that satisfies

' ae volume-concentration requirements 1.

At least once per month -

of Figure 3.4.1 and at all times when the standby liquid poison system is re-Doron concentration shall be quired to be operable the temperature determined.

In addition, the boron shall not be less than the solut ion temp-concentration shall be detemined emture presented in Figure 3.4.2.

In any time water or boron are added addition, the heat tracin;; on the pump or if the solution ter:perature drops suet!an lines shall 1o operable whenever below the limits specified by the e com t emperature I: less ti an the Figure 3.L.2.

solution temperature presented in hgure 3 4.2.

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EXHIBLT B (Continued) 30 LII41TIIU COUDITIONS FOR OPERATIOIl 4.0 SUR7E111 alice F1QUIFBEITIG 3

14 hen irradiated fuel is in the reactor vessel and reactor coolant temperature is 3

less than 212 F, all low pressure core and contairm.ent cooling subsystems may be inoperable provided no work is being done which has the potential for draining the reactor vessel.

H.

Extended thintenance Jhen it is detemined that maintenance to restore components or systems to an operable condition will last longer than the periods specified, m._p-r' dc^n11170 M.c c i

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Bases 3.6 and 4.6 - Continued:

he initial IWf Temperature of the vessel shell material opposite the reactor core region is O F.

he initial IUf temperature in the main dome flanges, and the shell and head material connecting to these flanges is 10 F, cnd elsewhere is 400F.

The design life of the reactor vessel is 40 years and the maximum fast neutron exposure at 40 years is calculeted to be 5.h x 1017 nyt.

The NDT temperature limit curve in Figure 4.6.1 uses the " worst case" curve of the FSAR to establish the NDT temperature shift and is therefore conservative. The expected IUf temperature shift for this vessel at 5 4 x 1017 nyt is expected to be O F.

Figure 4.6.1 also incorporates a 60UF factor of safety. This factor is based upon the require-ments of the !SE code and the considerations which resulted in these requirements.

Berefore, the specification provides for " worst case" data as well as 60 F of margin to provide assurance that g

operation in the non-ductile region will not occur.

i Le reactor vessel head flange and the vessel flange in combination with the doub)e "O" ring type seal i mgned to provide a leak-tight seal when bolted together. Vnen the vessel head is placed on the are res.:tcr veusel, only that portion of the head flande near the inside of the vessel rests on the vessel flange.

As the head bolts are replaced and tensioned, the vessel head is flexed slightly to bring together the entim contnet sitriaces adjacent to the "O" rings of the head and vessel flange.

Both the head and the headflange have an IUf temperature of 10 F, and they are not subject to any appreciable neutron radiation exposure. Eere-fore, the minimum vessel head and head flange temperature for bolting the head flange and vessel flange is established as 10 F + 60 F or 700F.

Numerous data am available relating integrated flux and the change in nil-ductility transition temperuture (NDPf) in various steels. Se most conservative data has been used in Specification 3 3 he integrated flux at the veasel wall is calculated from core physics data and will be measured using flux monitors installed ins?

the vessel.

Le measurements of the neutron flux at the vessel wall will be used to check and if nee-sary correct, the calculated data to detemine an accurate NUPf.

O In addition, vessel material samples will be located within the vessel to monitor the affect of neutron exposure on these materials. The samples include specimens of base metal, weld zone metal, heat affected zone metal, < nd standard cpecimens. These samples will receive neutron exrocure more rapidly than the vessel wall material and taerefore will lead the vessel in integrated neutron flux exposure. Rese samples will provide furtner assurr_nce that the shift in NDTT used in the specification is conservative. An analysis and report will be submitted to the AEC on all such surveillance specimens removed from the reactor vessels in accordance with Specification U ".

t These reports shall include the infomation specified in AS'D4 E-185-66, "Fecommended Pract i ces for Surveillance Tests on Structural thterials in Nuclear Beactor," and infomation obtained on the le'. 31 of integmted fast neutmn irradiation received by the specimens and actual vessel mnterial.

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Sections 1.0. A,1.0.B and 1.0.C concicting of pages 1 and 1A.

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Section 6.0 consicting of pages 192 throwh 220 (29 pages).

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IITTRODUCTION

'Bese Technical Specifications are prepared in accordance with the requirements of 10CFR50 36 and apply to the Ibnticello nuclear Generating Plant, Unit No.1.

The baces for these Specifications are included for information and understandability purposes.

1.0 DEFINITIONS The succeeding frequently used terms are explicitly defined co that a uniform interpretation of the Specifications may be achieved.

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A.

Abnormal Occurrence - An abnormal occurrence means the occurrence of any plant condition that results in:

1.

A cafety system setting less conservative than the limiting cetting established in the Tecturical Specifications.

2.

Violation of a limiting condition for operation established in the Technical Specifications.

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An uncontrolled or unplanned release or radioactive material from any plant system designed to act as a boundary for such material in an amount of significance with respect to limits prescribed in Technical Specifications.

h.

Failure of one or more components of an engineered safety feature or plant protection sys-tem that causes or threatens to cause ' lie feature or system to be incapable of perfoming its intended function.

0 5

Abnomal degmdation of one of the several boundaries designed to contain the radioactive

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materials resulting from the fission process.

6.

Uncontrolled or unanticipated cirinces in reactivity greater than 1% A k/k.

7 observed inadequacies in the implementation or administrative or procedural controls such that the inadequacy causes or threatens to cause the existence or development of an unsafe condition in connection with the operation of the plant.

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Conditions arising from natural or offsite man-trade events that affect or threaten to affect the safe operation of the plant.

B.

Alteration of the Reactor Core - The act of moving any component in the region above the core support plate, below the upper grid and within the chroud.

(No.nal operating func-tions cuch as control rod movement using the normal drive mechanism, tip scans, SRif and IRM detector movements, etc., are not to be considered core alterntions. )

C.

Hot Standby - llot Standby means operation with the reactor critical in the startup mcxle at a power level just cufficient to maintain reactor pressure and temperature.

O O

1.0-1A 1A

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.,5 6.0 ADMINISTFliTIVE CONTHOIS 6.1 Organization A.

The Plant Minnger has the overall full-time onsite responsibility for safe operation of the facility.

During periods when the Plant M1 nager is unavailable, he may delegate this responsibility to other qualified cupervisory personnel.

B.

The Northern States Power corporate organizational structure re3ating to the operation of this plant is shown in Figure 6.1.1.

C.

'Ihe minimum functional organization for operation of the plant shall be as shown in Figure 6.1.2.

The minimum shift complement for special plant conditions shall be as follows:

1.

The unit shutdown and the reactor contains fuel: a licensed senior reactor operator on site; a licensed reactor operator in the control room; and a plant operator on site.

2.

During cold startup, while shutting down the reactor and during recovery ftt>m any reactor trip: n minimum of tuo licensed opemtors in the control room.

3 Fuel handling operations: a licensed senior reactor operator shall be directly in charge of any fuel handling operations and shall have no other concurrent duties.

'4. Any fuel on site: an individual who meets the qualification of a health physics g

technician shall be on site at all times that there is fuel on site.

D.

Minimum qualifications, training, replacement training and retmining of plant personnel shall be in accordance with that stated in the Standard for Selection and Training of i

Personnel for Nuclear Power Plants, ANSI N18.1 - 1971. The minimur frequency of the m -

training program shall be every two years. The training program shall be under the di-rection of a designated member of the plant staff.

I 3

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PRESIDENT n

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t GROUP VICE PRESIDENT-i, POE R SUPPLY I

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VICE PRESIDENT-VICE PRESIDENT-DIRECIOR OF POWER PRODUCTION &

NUCLEAR SUPPORT

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ENGINEERING SYSTEM OPERATION SERVICES l

9 SAC Administration i

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l N^

MANAGER OF POWER PRODUCIION SAFETY AUDIT h

COMMITTEE (SAC)

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GENERAL SUPERINTENDENT l

NUCLEAR POWER PLANT j

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I PIMIT MANAGER Audit & Review g

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MONTICELLO 6

FIGURE 6.1.1 NSP CORPORATE ORGANIZATIONAL w

RELATIONSHIP TO ON-SITE OPERATING ORGANIZATION

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i SUPERI!UENDE!Tr-SUPERINTENDEIE-OPERATION &

PLANT ENGINEERING &

MAINTENANCE

• RADIATION PROTECTIO!P 3R0 j

QUALITY ENGItEER*

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MAINTENANCE OPEFATIONS RADIATION PIAtE ENGINEER-PLANT ENGINEER-SUPERVISOR

• SUPERVISOR

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PROTECTION TECIU4ICAL*

OPERATIONS

• ENGINEER

• SHIFT SB0 SUPERVISOR I/S t

MECllANICAL LEAD PLANT 1/S TECIINICAL ENGINEERS FOR TECHNICAL SUPPORT AND EQUIPMENT &

SUPPORT &

NUCLEAR ENGIfEER-FOR OPERATION, ELECTRICAL REACTOR OPERATOR RO TECllNICIANS ING, INSTRUMENTATION, MAINTENANCE, SUR-MAINTENANCE FOR RADIATION CONTROLS & COMPUTER, VEILIANCE, TESTING GROUP PIA!U EQUIPtfENT 1/S PROTECTION &

AND INSTRUMENT MEN AND RETRAINING

& REACTOR CIIEMISTRY OPERATOR RO ASSISTANT PLANT I/S EQUIPMENT OPERATOR PIANT ATTENDAITF I/S MONTICELID NUCLEAR GENEPATING PIAfE CODE:

- Key Supervisor Minimum Functional Organization For I/S

- one/ shift on-Site Operating Group RO

- Licensed Reactor Operator SRO

- Licensed Senior Reactor Operator FIGURE 6.1.2 6.1 194

i-6.2 Review and Audit 4

Organizational units for the review and audit of facility operations shall be constituted au have

[

i the responsibilities and authorities outlined below.

I I

l.. Safety Audtt Committee (SAC) l The SAC must: verify that operation of the plant is consistent with company policy and niles, l

approved operating procedures and operating license provisions, review important proposed

)

plant changes, tests and procedures; verify that unusual events are promptly investigated and corrected in a manner which reduces the probability of recurrence of such events; and detect trends which may not be apparent to a day-to-day observer.

l I

j Audits of selected aspects of plant operation shall be performed with a frequency commensurate with their safety significance and in such a manner as to assure that an audit of all nuclear safety related activities is completed within a period of two years.

Periodic review of the l

audit programs should be perfomed by the CAC at least twice a year to assure that such audits

{

are being accomplished la acconlance with requirements of Tech lical Specifications. The au-I dits shall be perfomed in accordance with appropriate written instructions or procedures and should include verification of ecepliance with internal rules, procedures (for example, norml, I

j of f-nonnal, er:ergency, operating, maintenance, surveillance, test and radiation control pro-j cedures and the emergency and security plans), regulations involving nuclear safety and opera-ting license provisions; training, qualification and perfomance of operating staff; and cor-rective actions following abnomal occurrences or unusual events. A representative portion of j

procedures and records of the activities perfomed during the audit period shall be audited and,in addition, observations of perfonnance of operating and maintenance activities shall be included. Written reports of such audits chall be reviewed at a schedule meeting of the SAC g

and by appropriate members of management including those having responsibility in the area T

audited.

Follow-up action, including reaudit of deficient areas, shall be taken when indicated.

1.

Membe rship i

i The SAC shall consist of at least five (5) persons.

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The SAC Chaimin chall be a USP ranngement representative appointed by the Group Vice President -Power Supply.

Other SAC members chall be appainted by the Group Vice Preci-dent-Power Supply or cuch other person as he may designate.

Tie Chairnan shall appoint a Vice Chai rman !

n the SAC mentecchip to act in hic absence.

c.

No more than two membe rc of the SAC shall be from i;roups holdlig line resp-onsibility for operation ur the plant.

d.

'ihe SAC tembers chould collectively ve the capability required to review problems 5

in the rollowing areas:

nuclear i ar r plant operations, nuclear engineering, chem-ist.rj and ntdiochemictry, r:etall argy, inctrumentation and control, radiological carety, mechanical and elect rical enginee ring, and othe r appropriate fie3 ds asso-ciated with the unique characteristics of the nuclear power plant involved. When the nat ure ur a particular problem dictates, special consultants will be utilized, at ne ce c ca ry, to provide expert ndvice to the SAC.

2.

M2eting Frequency The EAC crWI 1

~e t on call by the Chairman but not lecc frequently than twice a year.

3 Quorum A majority or the pennnent members, i ncluding the Chaimin or Vice Chainnan.

No more than a minority of the quorum chull te f rom groupn holding line responsibility for the operation of the plant.

4.

Eesponalbilitier - The following subjects chould be reported to and reviewed by th(. SAC:

a.

Proposed teste and experimente, und their reculto, when such testo or experimente may ec nctitute un unreviewed carety question, ns defir.ed in Section 50 59, Part 50, Title 10, wle of Federal henlationc.

L.

Ecalustit:m or prop;ced enarrec to procedures, equipment and cystems completed under pro iclunc f Cection $0.7), P: rt 50, Title 10, Code or i eleral Regulations to verify that ruch propoced changec do not conctitute unrevievel cafety questions.

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