ML20056D943

Forwards Preliminary Comments on TS Sys 80+ (Chapter 16 CESSAR-DC) Submitted in App K.Status Sheet Also Encl

ML20056D943
Person / Time
Site:	05200002
Issue date:	07/30/1993
From:	Wambach T Office of Nuclear Reactor Regulation
To:	Brinkman C ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
References
NUDOCS 9308190043
Download: ML20056D943 (35)

Text

l [ 2' July 30, 1993 Docket No.52-002 .

I Mr. C. B. Brinkman, Acting Director  :

Nuclear Systems Licensing  ;

ABB-Combustion Engineering, Inc.

1000 Prospect Hill Road Windsor, Connecticut 06095-0500 ,

t

SUBJECT:

PRELIMINARY COMMENTS ON TECHNICAL SPECIFICATIONS (TS's) SYSTEM 80+

(CHAPTER 16 CESSAR-DC)

Enclosed are the first set of comments by the staff regarding the TS's submitted in Amendment K to CESSAR-DC Chapter 16. A status sheet is also enclosed.

Sincerely, Odgind Siennd Ry-Thomas V. Wambach,- Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor. Regulation -

Enclosure:

As stated cc w/ enclosure: '

See next page DISTRIBUTION: .

tLDocket File PDST R/F DCrutchfield TWambach  ;

PDR MFranovich SMarruder TEssig >

JNWilson PShea JMoore, 15B18 GGrant, 17G21 ACRS (11) 0FC: LA:PDST PM:EDSTo ,/ (A)SC:PDST NAME: PShea'),9 TWbb$cbsg TEssig#b DATE: 07/3(i ~79307/J8/93 07/3t/93 0FFICIAL RECORD COPY:

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ABB-Combustion Engineering, Inc. Docket No.52-002 cc: Mr. C. B. Brinkman, Manager Washington Nuclear Operations ABB-Combustion Engineering, Inc.

12300 Twinbrook Parkway, Suite 330 Rockville, Maryland 20852 Mr. Stan Ritterbusch Nuclear Licensing ABB-Combustion Engineering 1000 Prospect Hill Road Post Office Box 500 ,

Windsor, Connecticut 06095-0500 Mr. Sterling Franks U.S. Department of Energy NE-42 Washington, D.C. 20585 Mr. Steve Goldberg Budget Examiner 725 17th Street, N.W.

Washington, D.C. 20503 t Mr. Raymond Ng 1776 Eye Street, N.W.

Suite 300 Washington, D.C. 20006 Joseph R. Egan, Esquire Shaw, Pittman, Potts & Trowbridge 2300 N Street, N.W.

Washington, D.C. - 20037-1128 Mr. Regis A. Matzie, Vice President Nuclear Systems Development  !

ABB-Combustion Engineering, Inc.

1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095-0500 r

9 .,

Please find below the status of the CE System 80+ Tech Specs Review:

DIVISION / BRANCH STATUS COMMENTS Instrumentation and i;ht See Enclosure 1, will receive additional Controls comments by 8/31/93.

Division of Reactor Complete See Enclosure 2 Inspection and Licensee Performance (DRIL) ,

Iluman Factors Assessment Complete See Enclosure 3 Mechanical Engineering, Complete See Enclosure 4 Material & Chemical Engr,

& Civil & Geosciences Reactor Systems Open Will provide comments by 8/15/90.

Containment Systems and Open Will pre ... e comments by 8/15/93 Sesere Accident  !

i ProWillstic Safety Open The PRA xylew for Tech Specs is Assas.nent ten nti#.y scheduled for completion in August,1993.

Radiation Protection Complete See Enclosure 5 Plant Systems Open Will provide comments by 8/15/93.

Electrical Open Will provide comments by 8/15/93.

Division of Radiation Complete CE did not deviate from the Standard ,

Safety and Safeguards Technical Specifications. Therefore, the Emergency Preparedness Branch found the CE 80+ Technical Specifications acceptable and has no comments.

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Enclosure 1 The following questions and comments result from the Instrumentation and Controls Branch (IIICB) review of the CE System 80+ TS.

RPS Process Instrumentation 3.3.1 Condition A and B The CE Standard TS has required actions and completion times for the loss of one or more functions as well as the loss of one or more RPS Instrumentation channels. The System 80+ TS addresses the loss of RPS channels only. Ilow does the System 80+ TS distinguish between a channel and a functional unit and is it possible to bypass a channel and a functional unit simultaneously? IIow will the System 80+ TS address the loss of a PLC within a channel?

Condition C & D The CE Standard (Digital) TS addresses the loss of one or more functions as well as the loss of one or more automatic bypass removal channels for the RPS Instrumentation. The System 80+ TS addresses the loss of one or more automatic bypass removal channels in LCO 3.3.1 but does not address the loss of functions. IIow does the System 80+ TS distinguish between channels and functional units and is it possible to bypass a channel and a functional unit simultaneously? IIow will the System 80+ TS address the loss of a PLC within a channel?

Logarithmic Power Level 3.3.4 Condition C Condition C for the System 80+ corresponds to Condition E in the CE' Standard (Digital) TS for Logarithmic Power Level. The System 80+ TS addresses the failure to meet actions and completion times for the previous conditions.

The requirement for this condition is that the facility be in mode 3 with-the RTCBs open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The CE Standard TS requires a completion <

time of I hour for the same condition. Explain the deviation from the CE

i Standard TS for this condition.

Engir.eered Safety Fea_ Lure Actuation System ESFAS) Instrumentation 3.3.10 Condition A and B The CE Standard TS has required actions and completion times for the loss of one or more functions as well as the loss of one or more channels. The  :

System 80+ TS addresses the loss of ESFAS Instrumentation channels but not the loss of functions. IIow does the CE System 80+ TS distinguish between channels and functional units and how is the loss of a PLC within a channel addressed?

Condition C and D The CE Standard TS has required actions and completion times for the loss of one or more functions as well as the loss of one or more automatic bypass removal channels for ESFAS Instrumentation. The System 80+ TS addresses  :

the loss of automatic bypass removal channels in LCO 3.3.10 but does not '

address the loss of functions. IIow does the System 80+ TS distinguish between a functional unit and a channel and how is the loss of a PLC within a channel addressed?

Condition E The CE Standard TS addresses the failure to complete the required actions and completion times of the above conditions. The required action is 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 4 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Why is this condition not addressed in the System 80+ TS for the ESFAS Instrumentation  :

Engineered Safety Featura Actuation System ESFAS) Logic 3.J.11 1

Condition C  !

Condition C of the CE Standard TS corresponds to Condition D of the System 80+ TS. The CE Standard TS requires that at least one contact in the  !

affected trip leg of the ESFAS logic be opened immediately on the loss of one or more functions and the loss of two initiation logic channels in the same trip i leg. The System 80+ TS addresses only the loss of the two initiation logic channels. Ilow does the System 80+ TS distinguish between a functional unit i

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'i and a channel and how is the loss of a PLC within a channel addressed? j 1

Condition D  ;

t Condition D in the CE Standard _TS corresponds to Condition F in the System  !

80+ TS. The CE Standard TS addresses the loss ot'one or more functions

-with one actuation logic; channel inoperable. The System 80+ TS. only. l addresses the loss of one or more actuation logic channels. How does the-  :

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. System 80+ TS distinguish between a channel and a functional unit and how is the loss of a PLC within a channel addressed?

i Remote Shutdown Instrumentation 3.3.13  ;

Condition A ,

j The CE Standard TS addresses the loss of one or more functions and the -l System 80+ TS addresses the loss of channels. How does the System 80+ TS '

distinguish between a channel and a functional unit and how is the loss'of a  ;

PLC within a channel addressed? j

.i Accident Afonitoring Instrumentation 3.3.14 ~

q Condition A i I

a The CE Standard TS addresses the loss of one or more functions with the loss  !

of a channel and requires that the channel be restored within 30 days.> The - j System 80+ TS do not address the loss of functions,' only the loss of channels.

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1 IIowever, the completion time is more conservative than the CE Standard TS,  ;

i requiring the channel to be restored within 7 days. How does the System 80+

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TS distinguish between a channel and a functional unit and how is the loss of . j

-a PLC within a channel addressed? l 1

Condition D _

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^l The CE Standard TS addresses the loss of two Hydrogen hionitoring Channels- -

^q and requires that at least one he restored to operable status in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This l condition is not addressed in the System 80+ TS for accident monitoring.

Explain the deviation from the CE Standard TS for this condition.-

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Endosure 2 ,

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CESSARinnm-  ;

16.15.1 5.5 REVIEWS AND AUDITS [

i Reviews and Audits 5.5 5.5 REVIEWS AND AUDITS

-r F---

Tbc licensee shall describe the method (s) established to conduct independent reviews and  ;

audits. The methods may take a range of forms acceptable to the NRC. Dese methods l amy in:lude creating an organistional unit or a standing or ad hoc committee, or assigning  ;

individuals capable of conducting these reviews and audits. When an individual performs a review function, a cross disciplinary review determination is necessary. If deemed necessary, su:b reviews shall be performed by the review personnel of the appropriate discipline. Individual reviewers shall not review their own work. Regardless of the method 1 used, the licensee shall specify the functions, organistional arrangement, responsibilities, .

j appropriate ANS!/ANS 3.11981 qualifications, and reporting requirements of each functional element or unit that contributes to these processes. ,

[ Reviews and audits of activities affecting plant safety have two distin:t elements. De first I

. element is the reviews performed by plant staff personnel to ensure that day to day activities J are conducted in a safe manner. These reviews are described in Section 5.5.1. The second -

element, described in Section 5.5.2, is the [offsite) reviews and audits of unit activities and I  !

programs affecting nuclear safety that are performed independent of the plant stafic The i ,

[offsite] reviews and audits should provide integration of the reviews and audits into a cohesive program that provides senior level utility management with an assessment of .'

facility operation and recommends actions to improve nuclear safety and plant reliability. It ' !

should include an assessment of the effectiveness of reviews conducted a: cording to Section ~ l K 5.5.1. ,

l 5.5.1 Plant Reviews

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The licensee shall describe provisions for plant reviews (organiution, reporting, records)

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and the appropriate ANSI /ANS standard for personnel qualification.

5.5.1.1 Functions ,

The [ plant review method specified in Specification 5.5.1) shall, as a minimum, - ,

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incorporate functions that:

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

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b. Recommend to the [ Plant Superintendent) approval or disapproval of items i considered under Specifications 5.5.1.2.s through 5.5.1.2.f prior to their implementation, except as provided in Specification 5.7.1.3; (continued) l

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P SYSTEM $0+ 5.0-10 ,

Amendment K -

16.15-10 October 30,1992

C r-I'* C A g GE$12N./ G M n CERTIFIC ATION Reviews and Audits 5.5 5.5.1.1 Functions (continued) -

c. Obtain approval from the [ Plant Superintendent], or his designee, in accordance with approved administrative procedures, for each proposed test or experiment and proposed changes and modifications to unit systems or equipment that affect nuclear safety prior to implementation;

d. Determine whether each item considered under Specifications 5.5.1.2.s it ough 5.5.1.2.e constitutes an unreviewed safety question as defined in ,

10 CFR 50.59; and l

e. Notify the [Vice President - Nuclear Operations] of any safety significant disagxement between the [re- crganization or individual specified in Specification 5.5.1] and the [hs Juperintendent] within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

However, the [ Plant Superintendent) shall have responsibility for resolution of such disagreements pursuant to Specification 5.1.1.

5.5.1.2 Responsibilities The [ plant review method specified in Specification 5.5.1] shall be used to conduct, as a minimum, reviews of the following:

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a. All proposed procedures required by Specification 5.7.1.1 and changes thereto;

b. All proposed programs required by Specif' cation 5.7.2 and changes thereto;

c. All proposed changes and modifications to unit systems or equipment that [

affect nuclear safety;

d. All proposed tests and experiments that affect nuclear safety; and ,

c. Review and documentation ofjudgment concerning prolonged operation with protection channels placed in bypass since the last [ plant review i

meeting] and the repair of these channels.

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f. All proposed changes to these Technical Specifications (TS), their Bases, [

and the Operating License.

(continued)

SYSTEM 80+ 5.0-11 Amendment K 16.15 October 30,1992 ,

CESSAR1lu% mon i f

i Reviews and Audits ~!

5.5 l 5.5.2 10ffsitel Review and Audit .

De licensee shall describe the provisions for reviews and audits independent of the plant's staff (organistion, reporting, and records) and the appropriate ANSI /ANS standards for 1 personnel qualifications. - These individuals may be located onsite or offsite provided  !

organistional independence from plant staff is maintained. He [te:hnical] review  !

responsibilities, Specification 5.5.2.4, shall in:lude several individuals located onsite. - l

- i 5.5.2.1 Functions i

De [offsite review and audit provisions specified in Specification 5.5.2) shall, as a (

minimum, incorporate the following functions that:

a. Advise the [Vice President - Nuclear Operations] on all matters related to l nuclear safety; .

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b. Advise the management of the audited organistion, and [its Corporate -l Management and Vice President - Nuclear Operations), of the audit results 2 as they relate to nuclear safety; i 5

c. Recommend to the management of the audited organintion, and its l

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management, any corrective action to improve nuclear safety and plant K

operation; and l i

d. Notify the [Vice President - Nuclear Operations) of any safety significant -l disagreement between the [ review organistion or individual specified in  :

Specification 5.5.2] and the [orgamintion or function being reviewed *- .~

within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 3 '

L 5.5.2.2 (Offsite] Review Responsibilities ne [ review method specified in Specification 5.5.2] shall be responsible for the review of: >

a. De safety evaluations for changes to pro:edares, equipment, or systems,  !

, and tests or experiments completed under the provisions of 10 CFR 50.59, j to verify that such actions do not constitute an unreviewed safety question -  !

as defined in 10 CFR 50.59; l i

(continued) j

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i SYSTEM 80+ 5.012  !

I Amendment K- .

-. N eMariunR ._- i

CESSAR innnena i

Reviews and Audits 5.5 A

5.5.2.2 [offsite) Review Responsibilities (continued)

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

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

d. Proposed changes to TS and the Operating Licenac;  !

e. Violations of codes, regulations, orders, license requirements, and internal procedures or instructions having nuclear safety significance;

f. All Licensee Event Reports required by 10 CFR 50.73;

g. Plant staff performance;

h. Indications of unanticipated deficiencies in any aspect of design or ,

operation of structures, systems, or com;ments that could affect nuclear safety;

i. Significant a:cidental, unplanned, or uncontrolled radioactive releases, K including corrective action to prevent recurrence;

j. Significant operating abnormalities or deviations from normal and expected performance of equip:nent that affect nu: lear safety; and j

k. The performance of the corrective action system.

i Reports or records of these reviews shall be forwarded to the [Vice President -  ;

Nuclear Operations) within 30 days following completion of the review.

5.5.2.3 Audit Responsibilities ,

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The audit responsibilities shall encompass:

a. The conformance of unit operation to provisions contained within the TS l and applicable license conditions; ,

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b. The training and qualifications of the unit staff; l

(continued)  ;

i SYSTEM 80+ 5.0-13  ;

Amendment K 16.15-13 October 30,1992 '  ;.

CESS AR nainumu  :

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Reviews and Audits 5.5 ,

5.5.2.3 Audit Responsibilities (continued)

c. The implementation of a!! programs required by Specification 5.7.2;

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

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

Nuclear Operations).

Reports or records of these audits shall be forwarded to the [Vice President -

Nu: lear Operations] within 30 days following completion cf the review.

5.5.2.4 [ Technical] Review Responsibilities l

The [ technical] review responsibilities shall encompass:

s. Plant opersting chara:teristics, NRC issuances, industry advisories,  ;

Licensee Event Reports, and other sources that may indicate areas for .,

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improving plant safety;

b. Plant operations, modifications. maintenance, and surveillance to verify independently that these activities are performed safely and correctly and that buman errors are reduced as much as practical;

c. Intemal and external operstional experience information that may indicate l areas for improving plant safety; and

d. Making detailed recommendations through the [Vice President - Nuclear ,

Operations] for revising procedures, equipment modifications or other j means ofimproving nuclear safety and plant reliability.

5.5.3 Records .

Written records of reviews and audits shall be maintained. As a minimum these records .,

shall include:

a. Results of the a:tivities condu:ted under the provisions of Section 5.5;  !

i (continued) 3 r

SYSTEM 80+ 5.014 t Amendment K  !

16.15-14 October 30,1992

CESSAR Ennneuc=

I Reviews and Audits I

5.5 5.5.3 Records (continued) f

b. ' Recommendations to the management of the organization bebg audited;

c. An assessment of the safety significance of the review or audit findings;

d. Recommended approval or disapproval ofitems considered under Specifications 5.5.1.2.s through 5.5.1.2.f; and

c. Determination whether each item considered under Specifications 5.5.1.2.s through 5.5.1.2.e constitutes an unreviewed safety question as defined in 10 CFR 50.59.

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t SYSTEM 80+ 5.0-15 Amendment K 16.15-15 October 30,1992

CESSAR n!!!nemOn P

16.15.1 5.7 PROCEDURES, PROGRAMS, AND MANUALS Procedures. Programs, and Manuals 5.7  ;

5.7 PROCEDURES, PROGRAMS, AND MANUALS l

5.7.1 Procedures 5.7.1.1 Scope Written procedures shall be established, implemented, and maintained covering the following a:tivities:

a. The applicable procedures recommended in Regulatory Guide 1.33. Revision 2 ,

Appendix A February 1978;

b. The emergency operating procedures required to implement the requirements of '

NUREG-0737 and to NUREG 0737, Supple:nent 1 as stated in Generic Letter 82-33; i

c. Security plan implementation;

d. Emergency plan implementation; i i

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e. Quality assurance for effluent and environmental monitoring; }

f. Fire Protection Progratn implementation; and l i

g. All programs specified in Specification 5.7.2.  ;

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h. Modification of core protection calculator (CPC) addressable constants. These <

procedures shall include provisions to ensure that sufficient margin is maintained in -

CPC type I addressable constants to avoid excessive operator intera: tion with CPCs j during reactor operation.

Modifications to the CPC software (including changes to algorithms and fuel cycle ,

specific data) shall be performed in accordance with the most recent version of

'CPC Protection Algorithm Software Change Procedure', CEN 39(A)-P which l has been determined to be app'icable to the facility. Additions or deletions to CPC addressable constants or changes to addressable constant software limit values shall not be implemented without prior NRC approval.

(continued) i I

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SYSTEM 80+ 5.0-17

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Amendment K 1

16.15-17 October 30,1992

bESSAREniinem I

Procedures, Programs, and Manuals 5.7 5.7.1.2 Review and Approval Ea:b procedure of Specifi:ation 5.7.1.1, and changes thereto, shall be revies ed in accordance with Specification 5.5.1, approved by the [ Plant Supuintendent) or his designee in accordan:e with approved administrative procedures prior to implementation and reviewed periodically as set forth in administrative procedures.

5.7.1.3 Temporary Changes Temporary changes to procedures of Specification 5.7.1 may be made provided:

a. The intent of the existing procedure is not altered;

b. De change is approved by two members of the plant management staff, at least one of whom holds a Senior Rea: tor Operator license on the unit affected; and

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

5.7.2 Prorrams and Manuals The following programs shall be established, implemented, and maintained.

5.7.2.1 Radiation Protection Program procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure.

5.7.2.2 Process Control Program (PCP)

The PCP shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and pa:kaging of solid i

radion:tive wastes will be a:complished to ensure complian:e with 10 CFR 20,10-CFR 61, and 10 CFR 71; state regulations; burial ground requirements; and other requirements governing the disposal of solid radioactive waste.

(continued) -

SYSTEM 80+

5.0-18 Amendment K 16.15-18 October 30,1992

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Enclosure 3 - ,

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HHFB C0lMENTS ON .

SECTION 5.0, " ADMINISTRATIVE CONTROLS" j

0F CESSAR (SYSTEM 80+) DESIGN CERTIFICATION. j AMENDMENT K page 4 l 5.2.2.e.4 "..

.(generic Letter Letter82-12)'..."

82-12)..." should read- '

..(Generic ,

5.3.2 page 8

....[ Regulatory Guide'T.81]..." should- read

...[ Regulatory Guipe 1.8])..."

... acceptable to the NRC staff) and 10 CFR 55, I 5.4.1 page 9  ;

and, for appropriate..." should read i

" ... cceptable to the NRC staff],10 CFR 50.120, '

0 CFR 55, and, for appropriate..."

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4 5 Enclosure 4 i

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CESSAR EnliflCADON i

MSSVs 3.7.1 TABLE 3.7.1-2 t MAIN STEAM SAFETY VALVE LIFT SETTINGS VALVE NUMBER ,

SG #1 SG #2 LIFT psig, v

$(d ]

v t ,

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[1] [1] [1200]

[2] [2] [1235]

[3] [3] [1260] ,

[4] [4] [1260]

[5] [5] [1260]

[1] [1] [1200]

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[2] [2] [1235]

[3] [3] [1260] l

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[4] [4] [1260]

[5] [5] [1260] ,

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SYSTEM 80+ 3.7-4 i

Amendment I i

16.10-4 Dece_mbeir 2_1m1_990 '}

• I CESSARinancu. l l

16A.10 B 3.7 FLANT SYFIEMS i 16A.10.1 B 3.7.1 MAIN FTIAM SAFETY VALVES l

- MSSVs B 3.7.1 B 3.7 FLAST SYSTEMS  :

B 3.7.1 Main Si== Safety Valves - b i

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. t BASES  ;

BACKGROUND' The Maia Steam' Safety Valves (MSSVs) mainly provide over. pressure l protection for ebe ===dary symem. In doing so, the MS$Vs also pecmde  !

protection against overpressursnag the rescior coolant pressure boundary by . .)

providing a beat sink for removal of energy from the Ranctor Coolant System 1 (RCS) if the preferred best sink, provided by the twA- and Carculanag i Water system, is not evadable.  ;

• j Five Maia Steam Safety Valves, (ten per steam generstor) are located on each 'i Main Steam line, outside Coor===aat, upstream of the Main Steam Isolation  :

Valves, as describd in CESSAR-DC Section 5.2.2 (Ref.1).~ The MSSVs* -  ;

rated capacity passa the full steam' flow at 1025 RATED THERMAL 1 j POWER (RTP)(100 + 25 for instrument error) with the valves full open. ,  :

This meets the requirements of the ASME_ Code (Ref. 2) as described in the -  !

er-pressure Protection Report, CESSAR-DC Appendix 5.A (Ref. 3).pe  !

f MSSV design includes maggered seapoints, as shown in Table 3.7.1 2. so that  !

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• • • only the number of valves needed will actusts. Staggered setpoints reduce l m T '"* g .rpt stmu",s.. *31o .a-the potential for valve charternas because of insufficient steam pressure to ij

[3,A.,.,.u

,,1 funy ope. au valves fou wing . iurbia.-re.c., i,ip. y The valve lift settings given in Table 3.7.1-2 meet 'abe requirements of l Section III of the ASME Code (Ref. 2). The etal relieving capacity for all' l twenty MS$Vs at 1105 of system design pressure (adjusted for a $0 psi .  !

pressure drop to valves inlet) is 19 E6 lbea/br plus accumulation.1This .

j capacity is less than the total rated capacity because the MSSVs operate at an inlet pressure below rated conditions ensuring that mesa generator pressure j does not saceed 1105 of design. At these same secondary pressure j eooditions, the total flow at 1025 oQ,817 Mw]t (RTP plus 17 Mwt 'l

. pump best input) ~ 17.46 E6 lbm/br]The ratio of this total steam flow to  !

the total capacity is 09.95. ]  :

The low pressure sospots MSSV,1200 psia. wu, rt to a zero power, j

-l loop everage temperature (T,y,) (secondary fluid esrurstion tempersture) of j f 566 F. The RCS T,,, must be above this temperature to open MSSVs. j i

(continued) ~ f t

-i SYSTEM 80+ B 3.7-1 i, i

Amendment I - i

'11 16A.10-1 December 21,1990 - j q

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CESSAREn%um

b. Provisions for safety-related snutbers in accordance with 1D CrR 50.5Sa. The only snutbers excluded from this requirement are f Anstalled on mensafety related systems and then only if their I.

f ailure or f ailure of the system on which they are installed.

wouldnothaveanadverseeffectonanysafety-relatedsystem.)j Procedures, Programs, and Manuals 5.7 I

5.7.2.11 Inservice Inspection Program his program provides controls for inservice inspection of ASME Code Class 1,2, and 3 components, including applicable supports. He program shallinclude the '

I following:

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\ a. Provisions that inservice inspection of ASME Code Class 1,2, and 3 1 components shall be performed in accordance with ASME Boiler and ,

IDtiffs k Pressure Vessel Code and Addenda,Section XI, as required by 10 CFRf

'\ 50.55a(g) t whers(ief has beenguestedTro6M pursuant) ~

6C_FR 50:55a(g)(b)(i)'and II)'CFR 60.55Ea)(37 h ,_.

g C-The provisions of SR 3.0.2 are applicable to the frequencies for performing inservice inspection activities; -

/ An inservice inspection program for piping identified in NRC Generic

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/d . Letter 88-01 in accordance with the NRC staff positions on schedule, methods, personnel, and sample expansion included in Generic Letter 88-01, or in accordance with alternate measures approved by the NRC staff; and .

i f' Nothing in the ASME Boiler and Pressure Vessel code shall be construed e A to supersede the requirements of any TS. K 5.7.2.12 Inservice Testing Program ,

His program provides controls for inservice testing off T 'b 2 T'"--

?,

<liish3 components including applicable supports. The program st.allinclude the

.ggkaN ,

Provisions that inservice testing offSME Codtf Clas~s'iXand Ipumpg a.kvalvesCass-sou@l be performed in accordance with Section XI of  ;

the ASME Boiler and Pressure Vessel Code and applicable Addenda as ,

required by 10 CFRggW-whers4KeT has'beentequested .-

i ffrom the Commissionyursuanf tcy'10 FR40.55afr)(6)(i) a'nd 40DFR (Ji alptovided'in GLB -0$( ,

4 (continued)

SYSTEM 80+ 5.0-24 +

Amendment K 16.15-24 October 30,1992

~

'CESSAREn&bmx __.

b. Frovisions for safety-related snubbers in accordance with 10 CrR SD.$5a. Safety-related snubbers included those installed on safety-related components and t. hose installed on nonsafety-related components if their failure or failure of the component on which-they are installed would have an adverse effect on any safety-r.let.d syst .

Procedures, Programs, and Manuals ^

5.7

( 5.7.2.121nservice Testing Program (continued) "

L 5M - _ -

f.

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

i ASME Boiler and Pressure Vessel Code and applicable Addenda Required frequencies for terminology for inservice testing performing inservice testing activities activities  ;

Weekly At least once per 7 days

- Monthly At least once per 31 days >

Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days ,

Every 9 months At least once per 276 days Yearly or annually At least once per 366 days g ,

Biennially or every 2 years At least once per 731 days

. He provisions of SR 3.0.2 are enplicable to the above required

. Frequencies for performing inservice testing activities; #

pl ne provisions of SR 3.0.3 are applicable to inservice testing activities; e, and Nothing in the ASME Boiler and Pressure Vessel Code shall be construed f to supersede the requirements of any TS.

J,  !

5.7.3.13 Steam Generstor (SG) Tube Surveillance This program provides controls for monitoring SG tube degradation. Each SG shall be demonstrated OPERABLE by its meeting the requirements of Specification  ;

5.7.2.11 and by an approved augmented inservice inspection program that includes  !

at least the following:  !

a. SG sample selection and inspection;

b. SG tube sample selection and inspection; (continued)

SYSTEM 80+ 5.0-25 1

Amendment K 16.15-25 October 30,1992 ,

. -, . ._ _..m.. . . . - - . - .

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4 Additional Comments  ;

1. The . tech'nical specification includes provisions for- surveillance of -

Reactor Shield Building in Section 3.6.8.2 but does not-Include similar. _i provisions for steel containment - vessel (SCV). The' technical i

specification should include surveillance requirements for SCV to verify - .:

its structural integrity by' performing a visual inspection of the exposed - -!

interior and exterior surfaces of the SCV. 3 a

2. The technical specification should address the operation' and' maintenance of the seismic instrumentation. In particular the technical.  !

1 specification should delineate the steps that will be taken in the event that the OBE level earthquake is exceeded during the lifetime ' operation- l of the plant.  ;

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Enclosure 5 .

t Facility Radiation Protection Section (FRPS) sub-enclosure 1 .

Radiation Measurement and- i

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-IIcalth Effects Section (RMIES) sub enclosure 2' i

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Sudo ENCLOSURE 1 FRPS COMMENTS

1. Section 5.5 (Reviews and Audits) of the CE System 80+ TSs states that the licensee shall describe the method (s) established to conduct independent reviews and audits. The TS goes on to state that the licensee "shall specify the functions, organizational arrangement, responsibilities,...of each functional element or unit" that will be used to conduct these independent reviews and audits. Section 5.5 should be amended to state that any organizational unit or group established to conduct these independent reviews and audits shall include a health physicist as a full time group member (see markup).

2. Section 5.9.1.2 of the CE System 80+ TS should be modified to reference the revised Part 20 (i.e., chonge 20.407 to 20.2206) (see markup).

3. Section 5.11 (High Radiation Area) of the CE System 80+ TSs should be modified to make reference to the revised Part 20 (which all licensees are required to comply with by January 1, 1994). The FRPS staff is currently in the process of rewriting this portion of the TSs to comply with and reference the revised Part 20. The most recent version of this writeup is attached.

Mt i 1%Lk  ;

CESSAREHL m t, wang i

16.15.1 5.5 REVIEWS AND AUDITS  !

Reviews and Audits 5.5 j 5.5 REVIEWS AND AUDITS The licensee shall describe the method (s) established to conduct independent reviews and audits. The methods may take a range of forms acceptable to the NRC. Dese methods (a include creating an organintional unit or a standing or ad hoc commitiee, or assigning K in ividuals capable of conducting these reviews and audits. When an individual performs a review function, a cross disciplinary review determination is necessary. If deemed necessary, such reviews shall be performed by the review personnel of the appropriate discipline. Individual reviewers shall not review thQ own work. Regardless of the method used, the licensee shall specify the functions, organizational arrangement, responsibilities, appropriate ANSI /ANS 3.1-1981 qualifications, and reporting requirements of each

• • f cs%:Q g ceb s aese functional ish'felement or unit that u e.de.t res.'e.>.s contributes f < d Ms to these s Le n Ntprocesses.

en L e*A tW*< a(Lysm adbMsv a s t

• y'4 C '- I foU M*e 8ReTewsYd'a'udits of aW.ities affecting plant safety have two distinct elements. He first i element is the reviews performed by plant staff personnel to ensure that day to day activities  !

are conducted in a safe manner. These reviews are described in Section 5.5.1. De second i element, described in Section 5.5.2, is the [offsite] reviews and audits of unit activities and programs affecting nuclear safety that are performed independent of the plant staff. The

[offsite] reviews and audits should provide integration of the reviews and audits into a cohesive prograrn that provides senior level utility management with an assessment of facility operation and recomraends actions to improve nuclear safety and plant reliability. It  :

should include an assessment of the effectiveness of reviews conducted according to Section K  :

5.5.1. i 5.5.1 Plant Reviews The licensee shall describe provisions for plant reviews (organization, reporting, records) and the appropriate ANSI /ANS standard for personnel qualification.

5.5.1.1 Functions

]

The [ plant review metbod specified in Specification 5.5.1) thall, as a minimum,  ;

incorporate functions that:  :

i

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

~

b. Recommend to the [ Plant Superintendent] approval or disapproval ofitems )

considered under Specifications 5.5.1.2.a through 5.5.1.2.f prior to their I I

implementation, except as provided in Specification 5.7.1.3-

-l (continued) i SYSTEM $0+ 5.0-10 l 1

I Amendment K 16.15-10 October 30,1992 i

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sn ,m4 pg'%-s -a, CESSAR Eminem -

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Reponing Requirements 5.9 5.9.1 Routine Remns (continued) .

5.9.1.2 Annual Reports r NOTE A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the sta*. ion.

Annual Reports covering the activities of the unit as described below for the .

previous calendar year shall be submitted by March 31 of each year. [He initial repon shall be submitted by March 31 of the year following initial criticality.] .,

Reports required on an annual basis include:

a. Occupational Radiation Exposure Repon The tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures > 100 mrem /yr and their associated man rem exposure according to work and job i functions (e.g., reactor operations and surveillance, inservice inspection, 20.1104 routine maintenance, special maintenance [ describe maintenance), waste K-rocessing, and refueling). His tabulation supplements the requirements

% of 10 CFR'-20atet. The dose assignments to various duty functions may ,

be estimated based od pocket dosimeter, thermoluminescent dosimeter

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

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

5.9.1.3 Annual Radiological Environmental Operating Repon

- ---- - NOTE --

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

(continued) l SYSTEM 80+ 5.032 ,

Amendment K 16.15-32 October 30,1992

^

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$\bQ $ME = , hcm b DRAFT-5.11 HIGH RADIATION AREAS As pro 71ded in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraphs 20.1601 (a) and (b) of 10 CFR Part 20:

5.11.1 Hiah Radiation Areas with Dose Rates not Exceedina 1.0 ren/ hour:*

A. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be breached only during periods of personnel entry or exit.

B. Access to, and activities in, each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area (s) and other appropriate radiation protection equipment and measures.

C. Individuals qualified in radiation protection procedures (e.g., health physics technicians) and -

personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are following plant radiation protection procedures for entry to, exit from, and work in such areas.

D. Each individual (whether alone or in a group) entering such an area shall possess:

(i) A radiation monitoring device that continuously displays radiation dose rates in the area

(" radiation monitoring and indicating device"); or (ii) A radiation monitoring device that continuously .

integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached (" alarming. dosimeter"), with an ,

appropriate alarm setpoint, or (iii)A radiation monitoring device that continuously .

transmits dose rate and cumulative dose to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the_ area, or (iv) A self-reading dosimeter and, (a) Be under the surveillance, as specified in the RWP, while in the area, of an Individual i at the work site, qualified in radiation ]

protection procedures, equipped with a  ;

1 ,

4

________...m __

DRAFT radiation monitoring and indicating device who is responsible for controlling personnel radiation exposure within the area, or (b) Be under the surveillance, as specified in the RWP, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area.

E. Entry into such areas shall be made only after dose rates in the area have been determined and entry 1 personnel are knowledgeable of them. ,

5.11.2 Hiah Radiation Areas with Dose Rates Greater than 1.0 rem / hour.

• but less than 500 rads / hour:**

l A. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided '

with a locked door or gate that prevents unauthorized entry, and in addition:

(i) All such door and gate keys shall be maintained under the administrative control of the shift '

foreman or the health physics supervisor on duty.

(ii) Doors and gates shall remain locked except during periods of personnel entry or exit.

B. Access to, and activities in. each such area shall be i controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the  :

immediate work area (s) and other appropriate radiation '

protection equipment and measures.

C. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are following plant radiation protection procedures for entry to, exit from, and work in such areas.

D. Each individual (whether alone or in a group) entering '

such an area shall possess:

(i) An alarming dosimeter with an appropriate alarm setpoint, or (ii) A radiation monitoring device that continuously '

transmits dose rate and cumulative dose to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means 2

m DRAFT to communicate with and control every individual in the area, or (iii)A self-reading dosimeter and, (a) Be under the surveillance, as specified in the RWP or equivalent, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring and indicating device who is responsible for controlling personnel exposure within the area, or (b) Be under tha surveillance, as specified in the RWP or equivalent, by means of closed circuit television, of. personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area.

E. Entry into such areas shall'be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them.

F. Such individual areas that are within a larger area that is controlled as a high radiation area, where no enclosure exists for purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate, but shall be barricaded, conspicuously posted as a high radiation area, and marked by a conspicuous flashing light activated at the area as a warning device which is clearly visible from all access points to the area..

• At 30 centimeters (12 inches) from the radiation source or from any surface penetrated by the radiation.

• • At 1 meter from the radiation source or from any surface penetrated by the radiation.

1 I

HRACE , S/26/93a 3

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CESSAR !!!Memox 4

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Procedures, Programs, and Manuals 5.7 5.7.2.2 Process Control Program (PCP) (continued) i Licensee initiated cheges to the PCP:

a. Shall be documented and records of reviews performed ' be retained.

This documentation shall contain:

1. Sufficient information to suppon the chege(s) and appropriate analyses or evaluationsjustifying the change (s); and

2. A determination that the chege(s) maintain the overall conformance of the solidified waste product to the caisting requirements of Federal, State. or other applicable regulations.

Shall be effective after review and acceptance by the (review method of

, b.

Specification 5.5.1] and the approval of the [ Plant Superintendent).  ;

5.7.2.3 Offsite Dose Calculation Manual (ODCM)

a. 'ne ODCM shall contain the methodology and pamneters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring i alarm and trip se. points, and in the conduct of the Radiological Environmental Monitoring Program; K ,

b. The ODCM shall also contain the Radioactive Ef!)nm Controls and

. Radiological Environmental Monitoring programs required by Speci5 cation 5.7.2, and descriptions of the information that shoul:! be included in the Annual Radiological Environmental Operating, and Semiannual Radioactive Effluent Release Reports required by Specification [5.9.1.3) and Specification [5.9.1.4).

Licensee initiated changes to the ODCM: 9cfe ',s'

a. Shall be documented and records of reviews'performea shall be retained, This documentation shall contain:

1. Sufficient information to support the chang (s) together with the appropriate analyses or evaluationsjustifying the change (s),

(continued)

SYSTEM 80+ 5.0-19 f

Amendment K 16.15 October 30,1992

i CESSAR nha - ~ ~ -

l a

Procedures, Programs and Manuals ,

5.7 I

5.7.2.3 Offsite Dose Calculation Manual (ODCM)(continued) 9:

2. A determination that the change (s) * .ain the levels of radioactive effluent control requind 10 CFR ,40 CFR 190,10 CFR 50.36s, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or  !

serpoint calculations.

b. Shall become effective after review and acceptance by the [ review method of Specification 5.5.1) and the approval of the [ Plant Superintendent).

c. Sha!! be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Semiannual Radion:tive Effluent Release Report for she period of the report in which

. any change in the ODCM was made. Ea:b change shall be identified by '

, markings in the margin of the affected pages, clearly indicating the ans of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

l 5.7.2.4 Primary Coolant Sources Outside Contamment his program provides controls to minimi:e leakage from those portions of systems , ,

outside containment that could contain highly radioactive fluids during a serious i transient or accident to levels as low as practicable. He systems include [the !.ow K  ;

Pressure Core Spray, High Pressure Core Sprsy, Residual Heat Removal, Ree tor Core Isolation Cooling, hydrogen recombiner, process sampling, and Standby Gas Treatment). He program sha!! include the following:  :

a. Preventive maintens sce and period: visualinspection requirements; and

b. Integrated *:d est requirements for ea:b system at refueling cycle ,

i intervals or less.

5.7.2.5 In Plant Radiation Monitoring I

his program provides controls to ensure the capability to accurately determine the

- airborne iodine concentration in vital areas under accident conditions. 71,is program shallinclude the following: ,

a. Tnining of personnel; (continued)

SYS7EM 80+ 5.0-20 ,

Amendment K-16.15-20 October 30,1992

_ mmcw' w=*~mma-mm- sa we Spym** $ w CESSARnnhmu Procedures, Programs, and Meuds 5.7 5.7.2.5 In Plant Radiation Monitoring (continued)

b. Procedures for monitoring; and c.

Provisions for maintenance of umpling and analysis equipment.

5.7.2.6 Post Accident Sampling his program provides controls that ensure the capability d to obtain an reactor coolant, radioactive gases, and particulates ind! plant gaseous emue containment stmosphere samples under accident conditions. He progra indude the following:

- a. Training of personnel; b.

Procedures for umplies and andysis; sad c.

Provisions for maintensm:e of umpling and andysis equipment.

g 5.7.2.7 Radioactive Emuent Controls Program his program conforms to 10 CFR 50.36a for themuents control of radioactiv and for maintesg the doses to members of the publicODCM, from radion:tive e as low as resonably achievable. De programs shall bei contained in the t shall be implemented by operating procedures, following elements:

a.

Limitations on the functional capability of radioa:tive liquid and gaseous monitoring instrumentation including surveillsnee tests and setpoint ,

determination in accordance with/0the snethodology in the ODCM; ts,es ++.t uncuko-hen s'&

b.

Limitations on the concentrations of radios: . e materid relened emuents to unrestri:ted s.reas, conforming to t&cFR4% Appendix B 7able% Column 2tb /0 M o74./00 /-JZ.0,d '/O /

2

c. Monitoring, emuents he-b umpling, and andysis of radioactive parameters in the ODCM; g gh

( (continued) h 3

5.0-21 i SYSTEM EO+

Amendment K October 30,1992 u

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Procedures, Programs, and Manuals 5.7 5.7.2.7 Radioactive Efiluent Controls Program (continued)

d. Limitations on the annual and quanctly doses or dose commitment to a member of the public from radioactive watedals in liquid effluents released from ca:b unit to unrestricted areas, conforming to 10 CFR 50, Appendix 1; e

t

e. Determination of cumulative and projected dose contributions from radion:tive effluents for the current calendar quaner and current calendar '

year in accordance with the methodology and parameters in the ODCM at least every 31 days;

f. Limitations on the functional capabiliry and use of the liquid and gaseous '

• effluent treatment systems to ensure that appropriate portions of these

. systems are used to reduce releases of radioa:tivity when the projected doses in a period of 31 days would caceed 2% of the guidelines for the '

annual dose or dose commitment, conforming to 10 CFR 50, Appendix I;

.J os '

i

g. Limitations on the dose .e resulting from radioactive material released in id & # 6*"j h m gaseous effluents to areas beyond the site boundary sonfcm.ing n i; du g,M3"q ra:0.4; + ?.Tn n.C+-:; ybe

-usor li,12mic 6 Ac.r$tto >q *. (s2 c. O.) + (3) ) K '

9 A teW bedu w tos 6 er .Mu h. Limitations on the annual and quarterly air doses resulting from noble '

'b gases released in gaseous effluents from eneb unit to areas beyond the site

4. c4eLa f nc 3 CC W'M boundary, conforming to 10 CFR 50, Appendix I; l

@.j Mel

i. Limitations on the annual and quanerly doses to a member of the public A) E ;,h-lal, from iodine 131, iodine-133, tritium, and all radionuclides in particulate ac,e Iu, Ed' %

w 6 a.it radic neJ A form with half lives > B days in gaseous effluents released from ca:b unit -

O' to areas beyond the site boundary, conforming to 10 CFR 50 Appendix I; in *) adit,dsk W w( Iso pcAR'

• and

• : us.s h ^

y a. den * *E Limitations on the annual dose or dose comitment to any member of the UPg.te rr.rcMF " j. public due to releases of radioactivity and to radiation from uranium fuel y a,. cycle sources, conforming to 40 CFR 190. ,

ceg 5.7.2.8 Radiologica! Environmental Monitoring Program This program is for monitoring the radiation and radionuclides in the environs of the plant. The program shall provide representative measurements of radioactivity '

in the higbest potential exposure pathways and verification of the a: curs y of the (continued)

SYSTEM 80+ 5.0-22  !

Amendment K 16.15-22 October 30,1992

n 2

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~_

CESSAR nanc.no l

l 1

Reporting Requirements j 5.9 5.9.1.3 Annud Radiological Environmental Operating Report (continued) i, The Annual Radiological Environmental Operating Report covering the operation of i

the unit during the previous cdendar year shall be submitted by May 15 of each j year. He report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the e porting period. He material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

He Annual Radiological Environmental Operating Repon shall include the results +

of analyses of all radiological environmental samples and of all environmenta!

rsdiation measurements taken during the period pursuant to the locations specified ,

' in the table and figures in the ODCM, as well as summarized and tabulated results

. of these analyses and measurements in the format of the table in the Radiologica!

Assessment Branch Technical Position, Revision 1, November 1979. [He report shall identify the thermoluminescent dosimeter (T1 D) results that represent -

collocated dosimeters in relation to the NRC TLD program and the exposure period associated with each result.) In the event that s.ome individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the mir. sing results. The missing data shall be submitted in a supplemectary report as soon as possible.

I 5.9.1.4 Semiannud Radioactive Effluent Release Repon i NOTE  ;

A single submittal may be made for a multiple unit station. The submittal should l combine sections common to all units at the station; however, for units with '

separste redwaste systems, the submittal shd! specify the releases of radioactive material from each unit.

$Au h. ocfn g lof.u m. %m. hue.u . %

He Semiannual Radioac vc Effluent Release Report covermg tbpperation of we l

unit during the previous, _ i cf +...a; ehall be submitted i' ~ ~ ^;-

' ;-- . He report shallinclude a summary of the l 1,-- 1 quantities of radioactive liquid and gaseous effluents and solid waste released from -

the unit. The material provided shs11 he consistent with the objectives outlined in j

j the ODMC and Process Control Program (PCP) and in conformance with 10 CFR  ;

50.36a and 10 CFR 50 Appendia 1,Section IV.B.1.

(continued) l

)

SYSTEM 80+ 5.033 l

4 e

Amendment E . _ _ _ _ - _ _ _ .