ML20136C640

Forwards Draft SER Re Tech Spec Changes & Draft Transmittal Ltr to Licensee.Proposed Amend to License NPF-1 Approved. SALP Input Also Encl

ML20136C640
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	12/20/1985
From:	Kirsch D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:	Thompson H Office of Nuclear Reactor Regulation
References
TAC-44625, TAC-45175, TAC-49636, TAC-49765, TAC-51292, TAC-52809, TAC-52971, TAC-53106, TAC-53808, TAC-54118, TAC-56709, TAC-56813, TAC-57001, TAC-57109, TAC-57290, TAC-59183, NUDOCS 8601030382
Download: ML20136C640 (33)

Text

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s DEC 2 01985 MEMORANDUM.FOR:

Hugh L. Thompson, Jr., Director Division of Pressurized Water Reactor (PWR)

Licensing - A i

FROM:

Dennis F. Kirsch, Acting Director Division of Reactor Safety and Projects

SUBJECT:

REGION V SAFETY EVALUATION REPORT FOR TROJAN NUCLEAR STATION MISC. TECH. SPEC. CHANGES (LCA 118)

TAC NO. 57290 Plant Name:

Trojan Docket Number:

50-344 Responsible Directorate:

PWR Project Directorate No. 3 Project Manager:

K. Johnston Review Status:

Complete Enclosed is the Region V Safety Evaluation Report addressing the subject licensing action and a draf t transmittal letter to the licensee.

Inasmuch as this action has been completed, TAC Number 57290 may be closed.

We have also enclosed the SALP input requested by NRR Ofrice Letter Number 44.

If you have any questions regarding this SER, please contact Jerry Zwetzig at FTS 463-3749.

Cri; k ) m :s Q R.J. Pgv Dennis F. Kirsch, Acting Director Division of Reactor Safety and Projects

Enclosures:

As Stated cc w/ enclosure:

K. Johnston, NRR-S. Varga, NRR-J. Carter, NRR-

.S. Richards, RV R. Dodds, RV D. Pereira, RV R.-Pate, RV G. Zwetzig, RV via 552J: ' Division of Licensing, NRR, Attn:

K. Johnston ID: Trojan TAC No. 57290 bec w/ enclosure:

Mr. Martin; Mr. Faulkenberry; Mr. Cook;_ Resident Inspector; Project. Inspector; RSB/ Document Control Desk (RIDS)

REGION V d

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K[LR[ffff [g' ZWETZIG PA H

12//7/85 12/n/85 12/d/85

\\il 8601030382 851220 PDR ADOCK 05000344

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s SALP INPUT'FOR COMPLETED SER Organization Preparing SALP Input: Region V Facility: Trojan Docket'No.: 50-344 Phase: Operating SER

Subject:

LCA 118 - Misc. Tech. Spec. Changes TAC No.: 57290 FUNCTIONAL AREA:

Licensing Activities Evaluation Criteria:

1.

Management Involvement in Assuring Quality The licensee's submittals reflected a management program where activities affecting quality were adequately controlled.

Rating: Category 2 2.

Approach to Resolution of Technical Issues from a Safety Standpoint With a few minor exceptions, the licensee's resolution of technical issues from a safety standpoint was very good.

In some instances the licensee requested technical specifications with requirements that go beyond the requirements of the Standard Technical Specifications.

Rating:

Category 1 3.

Responsiveness to NRC Initiatives The licensee was responsive to the staff's request to withdraw a certain portion of the original submittal.

Rating:

Category 2 4.

Reporting and Analysis of Reportable Events Not observed Rating: None 5.

Staffing (Including Management)

Not observed Rating: None 6.

Training and Qualification Effectiveness Based on the quality of the licensee's submittal, it appears the individuals preparing the submittal were adequately trained and qualfied.

Ratings:

Category 2

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t DRAFT Docket No. 50-344 l

Mr. Bart D. Withers

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4 Vice President Nuclear Portland General Electric Company i

121 S.-W. Salmon Street

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Portland,' Oregon 97204 i

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Dear Mr. Withers:

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Subject:

AMENDMENT NO.

TO FACILITY OPERATING LICENSE NO.'NPF-1 MISCELLANEOUS CHANGES TO THE TECHNICAL SPECIFICATIONS (LCA 118)

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4 The Commission has issued the enclosed Amendment No. _ to Facility Operating

- _ License No. NPF-1 for the Trojan Nuclear Station. This amendment revises the facility Technical Specifications in'respon_se-to your application dated

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March 12, 1985, as amended by your letter of. August ~22, 1985. One of the

. requests contained in your application, however,;has not been approved.

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The amendment revises a number of areas of the" technical specifications in-l l

order to clarify existing operability and surveillance requirements, to add.

new requirements or to correct errors that have been identified.

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A copy of the Safety Evaluation is also enclosed. The notice of issuance will l

be included in the Commission's next regular bi~-weekly Federal Register i

i notice.

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- Sincerely, l

3, -

a f

PWR Project Directorate No. 3

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Division of. Pressurized Water Reactor (PWR) Licensing

'A r

j Enc'losures:

1.

' Amendment No. _ to NPF-l'

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2.

Safety Evaluation i

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cc w/ enclosures:

j See next page

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9 DRAFT PORTLAND GENERAL ELECTRIC COMPANY THE CITY OF EUGENE, OREGON PACIFIC POWER AND LIGHT COMPANY DOCKET NO. 50-344 TROJAN NUCLEAR PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

License No. NPF-1 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Portland General Electric Company,,

et al., (the licensee) dated March 12, 1985, as revised by letter dated August 22, 1985, complies with the sta'ndards and requirements

'of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's Rules and Regulations set forth in 10 CFR Chapter I; B.

The facility will operate i conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

.C.

There_ is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or the health and safety of the public;-

E.'

The issuance of this amendment is in accordance.with 10 CFR Part 51 of the Commissior.'s regulations all applicable requirements have been satisfied.

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2.

Accordingly, the license is amended'by changes,to the TechnicaN E,

Specifications as indicated in'th'e attachment!to'this license amendment,

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and paragraph 2.C.(2) of Facilityf0pe,ratin~g Lic.ense;No. NPF-1 l's;hereby.

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amended,to read as follows:

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(2) Technical Specifications ~ -

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>8 The Technical Specificat' ions cont' ined.in Appendices A and B, as a

revised through Amendment Nog g_,_a,re hereby' incorporated in'the license. The licensee shall operate the ' facility in accordance with the Technical Specifications,'except'where otherwise stated int specific license' conditions.'.

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

FOR THE NUCLEAR REGULATORY COMMISSION PWR Project Directorate #3 Division of Pressurized Water Reactor (PWR) Licensing-A

Attachment:

Changes to the Technical Specifications Date of Issuance: October 4, 1985 t

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ATTACIDIENT TO LICENSE AMEND?'ENT NO.

TO FACILITY OPERATING LICENSE NO. NPF-1 DOCKET NO. 50-344 Revise Appendix A as follows:

Remove Pages Insert Pages 2-6 2-6 3/4 1-15 3/4 1-15 3/4 1-16 3/4 1-16' 3/4 1-20 3/4 1-20 3/4 2-2 3/4 2-2 3/4 4-2e 3/4 4-2e 3/4 5-4 3/4 5-4 3/4 7-14 3/4 7-14 3/4 7-16 3/4 7-16 B 3/4 1-1 B 3/4'l-1 B 3/4 1-2 B 3/4 1-2 B 3/4 1-3 B 3/4 1-3 B 3/4 1-4 B 3/4 1-4 B 3/4 2-2 B 3/4 2-2 B 3/4 4-2 B 3/4 4-2 B 3/4 5-1 B 3/4 5-1 6-3 6-3 t

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i TA8tE 2.2-1 (Continued) g o

[2 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

$iG FUNCTIONAL UNIT TRIP SETPOINT ALLOWA8LE VALUES

13. Steam Generator Water 2 5% of narrow range instrument 24% of narrow range instrument Level - Low-Low span - each steam generator span - each steam generator i

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14. Steam /Feedwater Flow

$ 40% of full steam flow at 5 42.5% of full steam flow at 1

Mismatch and Low Steam RATED THERMAL POWER coincident RATED THERMAL POWER coincident I

Generator Water Level with steam generator water level with steam generator water level t 25% of narrow range instrument 1 24% of narrow range instrument

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span - each steam generator span - each steam generator

15. Undervoltage - Reactor 2 68% each bus (8.48 kv) 1 67% each bus (8.35 kv) l Coolant Pumps i

16. Underfrequency - Reactor 2 57.5 Hz - each bus t 57.4 Hz - each bus m

!n Coolant Pumps i

17. Turbine Trip l

A.

Low Trip System 1 800 psig 2 700 psig Pressure 1

8.

Turbine Stop Valve 1 1% open 1 1% open Closure

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18.

Safety Injection Input Not Applicable Not Applicable.

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from ESF 3

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19. Reactor Coolant Pump Not Applicable Not Applicable j

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Breaker Position Trip i

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. REACTIVITY CONTROLS SYSTEMS BORATED WATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.7 As a minimum, one of the following borated water sources shall be OPERA 8LE:

4 a.

A boric acid storage system with:

l 1.

A minimum contained volume of 10,000 gallons, 2.

Between 7,000 and 7,700 ppe of boron, and j

3.

A minimum solution temperature of 65'F.

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The refueling water storage tank with:

1 1.

A minimum contained volume of 102.000 gallons, 2.

Between 2,000 and 2,500 ppm of boron, and i

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A minimum solution temperature of 37'F.

APPLICABILITY: NODES 5 AMD 6 4

l ACTION 1

!l With no borated water sources OPERA 8LE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one borated I

water source is restored to OPERA 8LE status.

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SURVEILLANCE REQUIRENENTS l

i 4.1.2.7 The above-required borated water source shall be demonstrated j

OPERABLE:

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At least once per 7 days by:

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1.

Verifying the boron concentration of the water,

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2.

Verifying the water level.of the tank, and 3.

Verifying the boric acid storage tank solution i

temperature when it is the source of borated water and the outside ambient air temperature is <37'F.

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TROJAN-UNIT 1 3/4 1-15 Amendment No. 58, 55, i

i' REACTIVITY CONTROLS SYSTENS 1

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80 RATED WATER SOURCES - OPERATING LINITING CONDITION FOR OPERATION i

3.1.2.8 Each of the following borated water sources shall be OPERA 8LE:

a.

A boric acid storage system with:

1.

A minimum contained volume of 15,900 gallons, l

2.

Between 7,000 and 7,700 ppe of boron, and l

3.

A minimum solution temperature of 65'F.

b.

The refueling water storage tank with:

1.

A minimum contained volume of 428,000 gallons of water, I

2.

Between 2000 and 2500 ppe of boron, and 3.

A minimum solution temperature of 37'F.

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APPLICABILITY: NODES 1, 2, 3 and 4.

ACTION:

a.

With the boric acid storage system inoperable, restore the storage system to OPERA 8LE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUTDOWN NARGIN equivalent to at least 1% ok/k at 200*F; restore the boric acid storage system to OPERA 8LE status within the next 7 days or be in l

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

b.

With the refueling water storage tank inoperable, restore the tank i

to OPERA 8LE status within one hour or be in at least NOT STANO8Y i

within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIRENENTS i

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4.1.2.8 Each borated water source shall be demonstrated OPERABLE:

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TROJAN-UNIT 1 3/4 1-16 Amendment No. 55, 4

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i REACTIVITY CONTROL SYSTENS

_PgilTION INDICATOR CHANNELS LIMITING CONDITION FOR OPERATION i

3.1. 3. 2 Control rod position indication system for control and shutdown l

rods and the demand position indication system shall be OPERABLE and capable of determining the control rod positions with !12 steps.

APPLICA8ILITY: MODES 1 and 2.

ACT,I_QN.:

3 a.

With a maximum of one rod position indicator per group inoperable l

either:

1.

Determine the position of the non-indicating rod (s) indirectly by the movable incore detectors at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and i

immediately after any motion of the non-indicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, or 2.

Reduce THERMAL POWER TO < 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

i b.

With a maximum of one demand position indicator per bank inoperable either:

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Verify that all rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of t

l each other at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or 2..

Reduce THERMAL POWER to < 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.3.2 Each rod position indicator shall be determined to be OPERABLE l

by verifying the demand position indication system and the rod position indication system agree within 10 steps at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except l

during time intervals when the Rod Position Deviation Monitor is inoperable.

l then compare the demand position indication system and the rod position j

indication system at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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TROJAN-UNIT 1 3/4 1-20 Amendment No. 70,

O POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued) b.

THERMAL POWER shall not be increased above 90% of RATED THERMAL POWER unless the indicated AFD is within the 15% target band and ACTION 2.a)1), above has been satisfied.

c.

THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD has not been outside of the 15%

target band for more than.1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty deviation cumulative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

t SURVEILLANCE REQUIREMENTS

'4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER j

by:

Monitoring the indicated AFD for each OPERABLE excore channel:

a.

1.

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2.

At least on;a per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the AFD Monitor Alarm to OPERABLE status, b.

Monitoring and logging the indicated AXIAL FLUX OIFFERENCE for each.0PERABLE excore channel at least once per 30 minutes l

when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable. The logged values of the indicated AXIAL FLUX OIFFERENCE shall be assumed to exist during the interval preceding each logging.

4.2.1.2 The indicated AFD shall be considered outside of its 15% target band when at least 2 of 4 or 2 of 3 OPERA 8LE excore channels are indicating the AFD to be outside the target band.

POWER OPERATION outside of the 5%

target band shall be accumulated on a time basis of l

i a.

One minute penalty deviation for each.one minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and,

b.

One-half minute penalty deviation for each one minute of POWER OPERATION outside of the target band at THERMAL POWER levels below 50% of RATED THERMAL ~ POWER.

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TROJAN-UNIT 1 3/4 2-2

,_-..__._..__,.____._._.,_.m

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

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS NORMAL OPERATION i

i LIMITING CONDITION FOR OPERATION 3.4.1.4 A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures 1 290*F unless:

1 a.

Another reactor coolant pump is running, or 1

b.

The secondary side temperature of each steam generator is less than 50*F above each of the.RCS cold leg temperatures a

j (ie,TSG < TC + 50*F), and overpressure protection systems are operable in accordance with Specification 3.4.9.3, or i

c.

A bubble has been established in the pressurizer with a 4

minimum vapor volume of 200 ft3 (895 pressurizer level).

i APPLICA81LITY: MODES 4 and 5.

ET.I.E:

If a reactor coolant pump is started and the above conditions are not met, a Spec *1al Report shall be prepared and submitted to the Coenission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the pump operation.

I SURVEILLANCE REQUIREMENTS 4.4.1.4 Not applicable, f

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f TROJAN-UNIT 1 3/4 4-2e Amendment No. 78,

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4

4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

I a.

At least once per 31 days by verifying that the following valves are in the indicated position with power to the operators removed:

Valve Number Valve Function Valve Position a.

NO 8806 a.

RWST Isolation a.

open*

b.

MO 8812 b.

RHR Suction a.

open*

c.

MO 8835 c.

SIS Cold Leg Injection c.

open*

d.

MO 8802-A d.

SIS Hot Leg Injection d.

closed l

e.

MO 8802-8 e.

SIS Hot Leg Injection e.

closed, t.

NO 8703 f.

RHR Hot Leg Discharge f.

closed-g.

MO 8809-A g.

RHR Cold Leg Discharge g.

open*

h.

NO 8809-8 h.

RHR Cold Leg Discharge h.

open*

1.

NO 8811-A 1.

Recir. Sump, RHR Suction 1.

closed *

j. NO 8811-8

j. Recir. Sump, RHR Suction

j. closed

• k.

NO 8813 k.

SI Pump Mini-flow isolation k.

open*

1.

MO 8814 1.

SI Pump Mini-flow isolation 1.

open*

b.

At least once* per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position.

c.

By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause i

restriction of the pump suctions during LOCA conditions.

This visual inspection shall be performed:

1.

For all accessible areas of the containment prior to establishing containment integrity, and 2.

Of the areas affected within containment at the completion of each containment entry when containment integrity is established.

• Power to be restored and valves operated from within control room for switch from injection to recirculation mode following LOCA.

TROJAN-UNIT 1 3/4 5-4 Amendment No. 74, l

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PLANT SYSTEMS l

COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3.2 At least one component cooling water train capable of supplying cooling water to equipment needed in MODES 5 and 6 shall be OPERABLE.

APPLICABILITY: MODES 5 and 6.

ACTION:

With less than the required number of component cooling water trains OPERABLE, declare supported equipment inoperable.

l SURVEILLANCE REQUIREMENTS 4.7.3.2 At least one component cooling water train shall be demonstrated OPERA 8LE by verifying that necessary components (the portion necessary to l

support equipment which is required to be OPERA 8LE in the applicable MODES above) are OPERABLE when tested pursuant to Specification 4.0.5 I

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l TROJAN-UNIT 1 3/4 7-14 Amendment No. 7#,

PLANT SYSTEMS SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4.2 At least-one service water train capable of supplying cooling water to equipment needed in MODES 5 and 6 shall be OPERABLE.

APPLICABILITY: MODES 5 and 6.

ACTION:

With less than the required number of service water trains OPERABLE, declare supported equipment inoperable.

l SURVEILLANCE REQUIREMENTS 4.7.4.2 At least one servce water train shall be demonstrated OPERABLE by verifying that necessary components (the portion necessary to support equipment which is required to be OPERABLE in the applicable MODES above) are OPERABLE when tested pursuant to Specification 4.0.5.

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TROJAN-UNIT 1 3/4 7-16 Amendment No. 7#,

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3/4.1 REACTIVITY CONTROL SYSTEMS BASES i

l 3/4.1.1 80 RATION CONTROL l

3/4.1.1.1 SHUTDOWN MARGIN l

A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made j

subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable.within i

acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

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SHUTDOWN MARGIN requirements vary throughout core life as a function l

of fuel depletion, RCS boron concentration, and RCS Tava. The most restrictive condition occurs at E0L, with Tave at no loid operating i

temperature, and is associated with a postulated steam line break accident' j

and resulting uncontrolled RCS cooldown.

In the analysis of this accident, i

a minimum SHUT 00WN MARGIN of 1.65 Ak/k is initially required to control the i

reactivity transient. Accordingly, the SHUT 00WN MARGIN requirement is based upon this limiting condition and is consistent with FSAR accident analysis assumptions. With Tava less than 200*F, the most restrictive reactivity transients resulting fFoe the postulated Boron 011ution Accident are such

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that a 1.65 Ak/k SHUT 00WN MARGIN is initially required.

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}/4.1.1.3 BORON DILUTION i

l A minimum flow rate of at least 3000 GPM provides adequate mixing, l

j prevents stratification and ensures that reactivity changes will be gradual i

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during boron ' concentration reductions in the Reactor Coolant System. A i

flow rate of at least 3000 GPM will circulate an equivalent Reactor Coolant System volume of 13,104 cubic feet in approximately 33 minutes. The i

reactivity change rate associated with boron reductions will therefore be within the capability for operator recognition and control.

i 1/4 The limitations on MTC are provided to ensure that the assumptions

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used in the accident and transient analyses remain valid through each j

j fuel cycle. The surveillance requirement for measurement of the MTC j

1 at the beginning and towards the end of each fuel cycle is adequate to l

l confirm the MTC value since this coefficient changes slowly due

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l TROJAN-UNIT I B 3/4 1-1 Amendment No. Si,

REACTIVITY CONTROLS SYSTEMS BASES 3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC) (Continued) principally to the reduction in RCS boron concentration associated with fuel burnup. The confirmation that the measured and appropriately compensated MTC value is within the allowable tolerance of the predicted value provides additional assurances that the coefficient will be maintained within its limits during intervals between measurement.

3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 551*F. This 11mit2 tion is required to ensure 1) the moderator temperature coefficien't is within its analyzed temperature range, 2) the pressurizer is capable of being in an OPERA 8LE status with a steam bubble 3) the reactor pressure vessel is above its minimum NOT temperature and 4) the protective instrumen-tation is within its normal operating range.

3/4.1.2 50 RATION SYSTEMS l

The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include 1) borated water sources, 2) charging pumps,

3) separate flow paths, 4) boric acid transfer pumps, 5) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above 200*F, a minimum of two separate and redundant boron injection systems are provided to ensure single func-tional capability in the event an assumed failure renders of the systems in-operable. Allowable out-of-service periods ensure that minor component l

repair or corrective action may be completed without undue risk to o.'erall i

facility safety from injection system failures during the repair period.

The boration capability of either system is sufficient to provide a SHUT 00WN MARGIN from all operating conditions fo 1.0% Ak/k af ter xenon decay and cooldown to 200*F. The maximum boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 14,418 gallons of 7000 ppe borated water from the boric acid storage tanks or 74,752 gallons of 2000 ppe borated water from the refueling water storage tank.

The required voluss for the boric acid storage tanks (two tanks) of l

14.418 gallons has been increased to a value greater than the minimum level indicating range of the storage tanks (741 gallons per tank) to 15,900 l

gallons.

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TROJAN-UNIT 1 8 3/4 1-2 Amendment No. 95

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REACTIVITY CONTROLS SYSTEMS l

BASES 3/4.1.2 80 RATION SYSTEMS (Continued)

With the RCS temperature below 200*F, one injection system is i

acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions pro-i h1 biting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.

1 The boron inventory in the RWST or the boric acid storage tanks is l

sufficient (a) to compensate for an inadvertent positive reactivity addition 1

j to the Reactor Coolant System of approximately 15 Ak/k while in MODE 5 at

[

200*F; and (b) to maintain a constant RCS reactivity while the temperature -

t l

in decreased from 200*F to 80*F.

In M00E 6, the boron inventory is suffi -

l cient to increase the boron concentration to compensate for an inadvertent positive reactivity addition of approximately 15 Ak/k while in the refuel-l ing mode. These conditions require 84g4 usable gallons of 7000-gpm borated I

water from the boric acid storage tanks or 23,432 usable gallons of 2000-gpm borated water from the refueling water storage tank.

The required volume for the boric acid storage tanks (two tanks) of l

84g4 gallons has been increased to a value greater than the minimum level indicating range of the storage tanks (741 gallons per tank) to gg76 gallons

(

(rounded to 10,000 gallons). The required RWST volume of 23,432 gallons must be increased to account for.nonusable volume due to tank geometry, i

i letdown and vortexing considerations (78,000 gallons), to 101,432 gallons 1

e

)

(rounded to 102,000 gallons),

f 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained. (2) the minimum SHUT 00WN MARGIN is main-tained, and (3) limit the potential effects of a rod ejection accident.

l' OPERASILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod i

1 alignment and insertion limits.

i i

The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that i

i the original criteria are met. Misalignment of a rod requires measurement i

j of peaking factors or a restriction of THERMAL POWER; either of these restrictions provides assurance of fuel rod integrity during continued operation. The reactivity work of a misaligned rod is limited for the remainder of the fuel cycle to prevent exceeding the assumptions used in j

the accident analysis for a rod ejection accident.

(

i TROJAN-UNIT 1 8 3/4 1-3 Amendment No. 58, 55, l

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l

REACTIVITY CONTROLS SYSTEMS BASES 3/4.1. 3 MOVABLE CONTROL ASSEM8 LIES (Continued)

The maximum rod drop time restriction is consistent with the assumed rod drop time used in the accident analyses. Measurement with Tavg >_550*F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

Control rod pos'itions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCO's are satisfied.

4 4

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t TRO.1AN-UNIT 1 83/41-4 AmendmentNo.58,//

March 3, 1982 I

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l POWER DISTRIBUTION LIMITS 8ASES Although it is intended that the plant will be operated with the AXIAL FLUX DIFFERENCE within the 15% target band about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion 4

will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels. This deviation will not affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached 6

l on a subsequent return to RATED THERMAL POWER (with the AFD within the l

target band) provided the time duration of the deviation is limited.

Accordingly, a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty deviation limit cumulative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is provided for operation outside of the target band but within the limits of Figure 3.2-1 while at THERMAL POWER levels between 50% & 90% of i

RATED THERMAL POWER.

For THERMAL POWER levels between 15% & 50% of rated THERMAL POWER, deviations of the AFD outside of the target band are less.

i significant. The penalty of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> actual time reflects this reduced '

significance.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process cc. iuter through the AFO Monitor Alarm. The computer determines the one minute average of each of the OPERABLE excore detector j

outputs and provides an alarm message immediately if the AFO for at least During. opera-l 2 of 4 or 2 of 3 OPERABLE encore channels are*outside the target band and the THERMAL POWER is greater than 90% of RATED THERMAL POWER.

tion at THERMAL POWER levels between 50% & 90% and 15% & 50% RATED THERMAL POWER, the computer outputs an alare message when the penalty deviation accumulates beyond the limits of I hour and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, respectively, i

Figure B 3/4 2-1 shows a typical monthly target band near the beginning

)

of core life.

l t

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TROJAN-UNIT 1 8 3/4 2-2 Amendment No.

l i

REACTOR COOLANT SYSTEM BASES j

The power operated relief valves (PORVs) operate to relieve RCS pres-sure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to proviie a positive shutoff I

capability should isolation of a relief valve be neiessary.

l 3/4.4.4 PRESSURIZER The requirement that 150 kw of pressurizer heaters and their associ-1 i

ated controls be capable of being supplied electrical power from an l

1 emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation l

at WOT STAN08Y. A minimum of 7 of the 23 kw heaters meets this requirement.

l 3/4.4.5 STEAN GENERATORS One OPERA 8LE steam generator provides suf ficient heat removal capa-i bility to remove decay heat af ter a reactor shutdown. The requirement for two OPERABLE steam generators, combined with other requirem6nts of the Limiting Conditions for Operation ensures adequate decay heat removal i

capabilities for RCS temperatures greater than 350*F if one steam gen-erator becomes inoperable due to single failure considerations. Below l

350*F, decay heat is removed by the RHR system.

I The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam j

generator tub ~es is based on a modification of Regulatory Guide 1.83, i

Revision 1.

Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the t

j event that there is evidence of mechanical damage or progressive degra-1 1

dation due to design, manufacturing errors, or inservice cenditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the 1

secondary coolant will be maintained within those parameter limits found to result in negligible corrosion of the steam generator tubes.

If the i

secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in stress corrosion cracking.

The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant TROJAN-UNIT 1 8 3/4 4-2 Amendment No. If,

3/4.5 EMER6ENCY CORE COOLING SYSTEMS (ECCS)

BASES I

3/4.5.1 ACCUMULATORS The OPERA 8ILITY of each safety injection system accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure j

falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe j

ruptures.

The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis 4

]

are met.

l The limit of one hour for operation with an inoperable accumulator minimizes the time exposure of the' plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures.

l 3/4.5.2 and 3/4.5.3.1 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration.

4 I

Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures l

within acceptable limits for all postulated break sizes ranging from the i

double ended break of the largest RCS cold leg pipe downward.

In addition, j

each ECCS subsystem provides long term core cooling capability in the recirculation mode during the accident recovery period.

With the RCS temperature below 350*F, one OPERA 8LE ECCS subsystem is acceptable without a single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling i

requirements.

The Surveillance Requirements provided to ensure OPERASILITY of each component ensures that at a minimum, the assumptions used in the safety l

analyses are met and that subsystem OPERASILITY is maintained. Surve11-i lance requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to:

(1) pre-vent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split j

between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow i

to all injection points equal to or above that assumed in the ECCS-LOCA analyses.

TRO,1AN-UNIT 1 8 3/4'5-1 Amendment No. 32, i

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.

TO FACILITY OPERATING LICENSE NO. NPF-1 PORTLAND GENERAL ELECTRIC COMPANY THE CITY OF EUGENE, OREGON PACIFIC POWER AND LIGHT COMPANY TROJAN NUCLEAR PLANT DOCKET NO. 50-344 I.

INTRODUCTION A.

DESCRIPTION OF PROPOSED ACTION The proposed action would amend various sections of Appendix A of the Technical Specifications for the Trojan Nuclear Plant (the facility) to correct or clarify the existing specifications and facilitate their use by the operating staff.

A Notice of Consideration of Issuance of Amendment to License and Proposed No Significant Hazards Consideration Determination and Opportunity for Hearing related to the requested action was published in the Federal Register on No j

comments or requests for hearing were received.

B.

. BACKGROUND INFORMATION By letter dated March 12, 1985, the Portland General Electric Company (the licensee) submitted License Change Application 118.

This application contained proposed revisions to various sections of the technical specifications including: Containment Integrity, Reactor Trip Instrumentation Setpoints, Borated Water Sources, Reactivity Control Systems, Power' Distributions Limits, etc. The proposed changes are intended to correct errors in the present specifications, clarify the meaning of the specifications or make the specifications easier to use by restating operating limits in terms of equivalent parameters that are readily monitored by the plant operators.

As a result of discussions with the staff, the licensee, by letter dated August 22, 1985,- withdrew that portion of the present application which, if implemented, would have changed the definition of Containment Integrity (Specification 1.8).

C.

SCOPE OF REVIEW i

This review has considered only the changes in the facility a

technical specifications requested by the licensee.

It has not considered those portions of the technical specifications for which changes were not requested.

In performing this review we have considered whether the changes would reduce any of the operational or administrative requirements implemented at the facility. Whether or not such a reduction was proposed, we have evaluated whether the change would:

(1) increase the probability or consequences of accidents considered in the FSAR, (2) create the possibility of an l

c.

2 l

accident not considered in the.JSAR, or (3) reduce the margin of safety as defined in the basis for any technical specification.

II.

EVALUATION a.

Table 2.2-1, Item 15.

This item specifies the required Trip

~

Setpoint and Allowable Value for undervoltage of the power supply for the reactor coolant pumps. The specification presently requires the Trio Setpoint to be equal to or greater than 68% for each bus and the Allowable Value to be equal to or greater than 67% for each bus. The licensee proposes to revise the specification by adding the specific voltage value corresponding to these percentage values.

Specifically, the licensee proposes to add a value of 8.48 kv in parentheses, following the 68% specification and a value of 8.35 kv in parentheses, following the 67% specification. The licensee states these voltage values are based on the bus design voltage of 12.47 kv as.specified in Updated FSAR Section 8.3.

We have confirmed this is the design voltage of the bus used to power the reactor coolant pumps, as stated in Section 8.3 of the Updated FSAR, and we have confirmed the correctness of the proposed voltage values. We also note the Standard Technical Specifications for Westinghouse Pressurized Water Reactors (STS), NUREG-0452, Revision 4, express these limits in terms of voltage, rather than as a percent of design voltage. Accordingly, we conclude this is an

[

editorial change, made for the purpose of clarification, and is acceptable.

b.

Specification 3.1.2.7.

This specification sets forth Limiting Conditions for Operation (LCOs) for Borated Water Sources when the reactor is in Operational Modes 5 and 6.

The licensee proposes to reword paragraph 3.1.2.7.a to refer to a " boric acid storage system" rather than the present reference to a " boric acid storage tank".

The licensee states this makes the specification consistent with the STS, which address a "boration system" rather than a single tank.

i Because the facility is equipped with two Boric Acid tanks supplying j

a common header, and because the change is consistent with the STS, we conclude this is an editorial change and is acceptable.

The licensee also proposes to increase the required minimum contained volume of Boric Acid from 9,235 gallons to 10,000 gallons.

l 1

The license states this is necessary to account for the 741 gallon minimum indication level of the second. tank when this specification is revised to cover both tanks.

(We note this specification was revised by Amendment 95 to account for the minimum indication level of one tank). With this addition, the resulting value would be 9,976 gallons; however, the licensee has rounded this upward to 10,000 gallons.

Because the proposed revision is more restrictive than the present specification, in order to account for the minimum indication level of a second storage tank, we conclude the proposed revision is acceptable.

The licensee has proposed revisions to the Bases for this specification to reflect the proposed changes described above. Our i

review indicates the proposed revisions are appropriate for the i

o 2.

a......

l 3

proposed changes in the specifications, and do not reduce the present margin of safety. Accordingly, we' conclude the proposed chang <s in the Bases for this specification are acceptable. We l

.have also corrected a typographical error in the licensee's l

submittal.

c.

Specification 3.1.2.8. 'This specification sets forth LCOs for l

Borated Water Sources when the ~ reactor is in operational Modes 1 through 4..The, licensee proposes to revise this specification by increasing the volume 'of borated water that must be available~ in the l

Boric Acid storage system,from 15,159 gallons to 15,900 gallons. As in the case of specification,3.1.2.7, this increase is proposed to

(

• account for the minimum indication level of the second storage tank.

Because the. proposed revision.is more' restrictive than the present specification, in~ order to account for this minimum indication level in a second storage tank, we conclude the proposed revision is c

acceptable.

As in the case.of specificStion 3.1.2.7, the licensee has proposed changes to the Bases which correspond to the proposed revisions to l

this specification.

We~ find the proposed revised Bases appropriate and acceptable.

j d.

Specification 3/4.1.3.2.

This specification tiets'forth operability requirements for the control rod position indicating systems. The licensee proposes to revise the wording of this specification to eliminate reference to rod position indicator " channels".

In support of this change, the licensee notes the rod position indicating system for each control rod consists of two position data channels, A and B, either of which alone can measure rod position within the required accuracy.

Because of this design, the use of the term " channel" in the present specification, lends itself to mis-interpretation. For example, one of the data channels for a given. control rod could be inoperable, but the position indication-system for that rod could still be operable. Nevertheless, because-of the present wording, such a condition might be interpreted as a position indicator channel being inoperable.

Our review of the STS indicates the revised wording proposed by the licensee conforms more closely to the STS wording. We also note the STS requires all rod position indicating systects to be operable, but does nct require the individual Data Channels comprising the system to be operable if that does not affect system operability.

Based on the above, we conclude the proposed revision clarifies the i

requirement of the specification in conformanc.e with the intent as i

expressed in the STS. Accordingly, we conclude the proposed change l

is acceptable, i

c.

Specification 4.2.1.1.b.

This specification sets forth the requirements for monitoring the reactor Axial Flux Difference at times when the Axial Flux Difference Monitor Alarm b inoperable.

The specification presently requires the monitoring,and logging of the Axial flux Difference for each operable excore flux channel at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the Axial Flux Difference Monitor Alarm is inoperable, and once per 30 minutes thereaf ter.

i'

=

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The licensee proposes to revise the specification to require monitoring the Axial Flux Difference every 30 minutes whenever the Alarm is inoperable; i.e. eliminate the one hour allowance during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The licensee notes this change is more restrictive, but requests it nevertheless, in the belief it will enhance safety and preclude potential violations of the technical specifications. This latter consideration derives from specification 3.2.1.a.2(a)(1), which limits operation with the Axial Flux Difference outside the Target Band to one hour cumulative in the preceding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. By monitoring the Axial Flux Difference more frequently, the licensee expects an improved capability to conform to this LCO.

The staff has considered this proposed revision and concludes it is more restrictive than present specifications with respect to the requirements for monitoring and logging the Axial Flux Difference, and does not reduce any other current requirements. Accordingly, we find the proposed revision acceptable, f.

Specification 3.4.1.4.

This specification sets forth limits on the startup and operation of reactor coolant pumps. The limits are imposed to prevent low temperature over pressurization events. The licensee proposes to revise the wording of sub paragraph b of this specification from:

"The secondary side temperature of each steam generator is less than 50*F greater than each of the RCS cold Icg temperatures...."

to:

"The secondary side temperature of each steam generator is less than 50 F above each of the RCS cold leg temperatures (ie TSG <

TC + 50 F)...."

The licensee states this proposed revision, including the equation, is intended to clarify the meaning of the specification.

Our review indicates the proposed revision does clarify the requirement and does not reduce any of the present requirements. Accordingly, we conclude this is an editorial change and is acceptable.

g.

Specification 4.5.2.a.

This specification lists a number of valves associated with the operation of the Emergency Core Cooling System (ECCS) and states the valve position must be verified to be correct each 31 days. The licensee proposes to' change the number of the salve appearing as Item d in this list- (SIS Hot Leg Injection) from MO 8002-A to MO 8802-A. The licensee states the present number is a

(

typographical error. We have reviewed,the applicable facility i

drawing (M-206) and confirmed the designations of'the SIS Hot Leg Injection valves are H0 8802-A and MO 8802-B.: Accordingly, we agcee f

that the present designation is incorrect'and the proposed revision 3

(

is appropriate and acceptable.

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b.

Specification 4.5.3.1.

This specification defines the surveillance requirements for the ECCS when the reactor is in the Hot Shutdown Operating Mode (350*F > T,'Ih=? 1 be demonstrated OPERABLE per the

>200 F).

At present this specification states "The ECCS subsystem applicable Surveillance Requirements of 4.5.2."

The licensee proposes to revise this specification to explicitly state which surveillance requirements (of 4.5.2) are applicable in this Mode.

Specification 4.5.2 lists nine areas of Surveillance Requirements --

designated 'a' through 'i'. Among these areas, the licensee proposes the following be defined as applicable to specification 4.5.3.1:

c, d.1, d.2, d.3.a, e.1, e.2.a, e.2.c, e.3, f.1, g.1, h, i.1, and i.3.

The surveillances the licensee' considers not applicable to this specification are discussed below.

As for areas 'a'.and

'b', which the licensee proposes to define as "not applicable", these surveillances involve periodic verification of the valve alignment necessary for automatic initiation of the ECCS. The licensee states these surveillances are not ne'cessary for the Mode covered by this specification (Mode 4) because the existing Limiting Conditio,ns for Operation (LCOs) for this Mode explicitly permit manual valve realignment. Our review,of the existing specifications confirms the LCOs'for Mode 4 permit manual realignment of ECCS valves. Accordingly, we agree the surveillances necessary to provide an OPERABLE flow path for auto-initiation of the ECCS in Modes 1, 2 and 3 are not required in Mode 4.

We also note, however, Mode 4 is a transitory condition usually lasting no more than a few hours during, plant startup cnd shutdown. The surveillances in question, however, are only' required once every 31 days (125%). Thus, since the' plant is normally in Mode 4 only for short periods of time, we see no need to explicitly waive the requirements for these surveillances while in Mode 4.

Further, we believe such an explicit waiver could be misinterpreted to justify not performing a surveillance that was due because the plant was in Mode 4.

Finally, the present specification (which follows the wording of the STS) allows the licensee to determine which surveillance requirements are applicable in Mode 4, and these determinations are subject to review by the NRC inspection program.

Based on these considerations, we conclude the licensee has not demonstrated a significant need for this proposed revision, and that i

the proposed revision could confuse, rather than clarify, present requirements. Accordingly, we conclude this proposed revision is not acesptable and have not included it in this amendment.

i i

i As for areas d.3.b, d.3.c, e.2.b, f.2, g.2, g.3 and i.2;-these surveillances all refer to valves or pumps that are part of the l

Safety Injection System. The licensee proposes to delete these requirements from this specification on the basis the technical-specifications do not require the Safety Injection System to be l

OPERABII in Mode 4.

l li

s 6

Regarding paragraphs d.3.b and d.3.c, these items require the position of certain ECCS manual throttle valves in the Safety Injection System be verified to be correct at least once every 18 months. As discussed above, because Mode 4 is typically a transitory condition, we question the need for such a waiver relative to a surveillance that is only required at 18 month (1 25%)

intervals.

Indeed, the only circumstance when such relief would be meaningful would be if the reactor were to remain in Mode 4 for a period in excess of 18 months.

Even then, however, upon leaving Mode 4 the licensee would be obliged to promptly perform the surveillance as described under specification 4.5.2.

Accordingly, because the licensee has not demonstrated a significant need for this change, and because the proposed revision could confuse rather than clarify the applicability of surveillance requirements, we do not find the portion of the proposed revision that would delete paragraphs d.3.b and d.3.c acceptable, and have not included it in this amendment.

Regarding paragraph e.2.b, this specification requires the testing of each Safety Injection Pump at least once per 18 months (125%),

during shutdown, to verify it starts automatically upon receipt of a safety injection signal. As with the case discussed above, we find this to be an important long-term pe' iodic surveillance that is not r

meaningfully related to transitory Operating Modes. Further, the fact the testing is to be performed when the reactor is shutdown (normally meaning Mode 5 - Cold Shutdown, or Mode 6 - Refueling),

suggests the surveillance is not performed in Mode 4.

The proposed revision, therefore, appears to have little or no meaning or significance.

In addition, as discussed ~above, the adoption of this revision could confuse, rather than clarify;the intent of the

~

specification. Accordingly, we do not find the portion of the proposed revision that would delete paragraph e.2.b acceptable, and

~

have not included it in this amendment.

Regarding paragraph f.2, this specification

• requires the' flow characteristics of Safety Injection cold leg piping be determined and adjusted during shutdown, following completion of modifications that would alter the flow characteristics of this piping.

It is the staff's view this is a clearly prudent requirement that is not altered by the fact the plant does or does not require certain equipment during a particular Operating Mode. Accordingly, we find the licensee has not provided sufficient justification for waiver of this requirement during Mode 4 and conclude this portion of the proposed revision is not acceptable. This proposed revision, therefore, is not included in the present amendment.

Regarding paragraphs g.2 and g.3, these items require verification of the correct positioning of specified ECCS manual throttle valves following stroking and or maintenance of the valves, when the associated ECCS system is required to be OPERABLE.

Incorporation of the proposed revision to the technical specifications would waive this verification requirement for the Safety Injection System when-the reactor was in Mode 4.

The licensee's basis for requesting this

-change is that this system is not required to be OPERABLE in Mode 4.

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Since the present specification already makes allowance for Operability requirements, we conclude this proposed revision would duplicate existing provisions and-is, therefore,-unnecessary.

Because the licensee has.provided no other basis for this proposed.

revision, we have not included-it in the present amendment.

Regarding paragraph i.2, this item requires the licensee to verify the Safety Injection Pump meets or exceeds a specified minimum discharge pressure when tested pursuant to Specification 4.0.5 (In-Service Testing). The proposed revision would waive this requirement when the reactor was in Mode 4, based on the fact the Safety Injection System is not required to be OPERABLE in Mode 4.

In-Service Testing is required by the facility Technical Specifications and 10 CFR 50.55a (g), and is not conditioned on Operational Modes except where a written request for relief, based on impracticality, has granted.

Inasmuch as no relief based on impracticality has been requested in this case, we conclude the proposed revision is not acceptable and it is not included in the present amendment.

i.

Specifications 3.7.3.2 and 4.7.3.2.

These are new technical specifications proposed to be added by the licensee. The proposed specifications address the LCOs and Surveillance Requirements applicable to the Component Cooling Water (CCW) system when the i

reactor is in Operating Modes 5 and 6.

In requesting these additions, the licensee notes that although the present LCOs for the CCW system are applicable only in Modes 1-4, the CCW system is also needed in Modes 5 and 6 to support systems required to be OPERABLE in these Modes, e.g. the Residual Heat-Removal Pumps and Heat.

Exchangers. The licensee also notes the definition of OPERABLE includes the requirement that supporting systems, such as cooling water systems, also be OPERABLE, but that this requirement is not addressed explicitly in the technical specifications.

To remedy this condition,~the licensee has proposed these additional technical specifications. The proposed Limiting Condition for Operation would require the operability in Modes 5 and 6 of at least one component cooling water train capable of supplying cooling water to equipment required to be OPERABLE in Modes 5 and 6.

The associated ACTION statement would be that if this LCO could not be met, the supported equipment would be required to be declared INOPERABLE. As a surveillance requirement, the licensee proposes the portion of the system necessary to support equipment required to be OPERABLE in Modes 5 and 6, be verified OPERABLE by testing in accordance with Specification 4.0.5 (In-Service Testing).

Based on our review of the proposed additions to the technical specifications, we find they provide clarification of, but do not reduce any present requirements. Accordingly, we conclude the incorporation of the proposed additional specifications will not reduce the safety of operations,, and such incorporation is acceptable.

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Specifications 3.7.4.2 and 4.7.4.2.

These are new technical specifications proposed to be added by the licensee to address the LCOs and Surveillance Requirements. applicable to the Service Water system when the reactor is in Operating Modes 5 and 6.

As with the CCW system discussed above, the licensee notes that although the present LCOs for the Service Water system are applicable only in Modes 1-4, this system is also needed in Modes 5 and 6 to support systems required to be OPERABLE in these Modes, e.g. the CCW system and Diesel Generators.

Accordingly, the licensee has proposed specifications for the Service Water system that are basically similar to those proposed for the Component Cooling Water system.' For the same reasons stated for that system, we find the additional specifications proposed by the licensee for the Service Water system to be acceptable, k.

Basis - Specification 3/4.1.1.4.

This specification sets forth LCOs and Surveillance Requirements for the Moderator Temperature Coefficient (MTC). The licensee proposes to revise the Basis for this specification to be consistent with the actual requirements of the specification. At present, the Basis refers to "... measurement of the HTC at the beginning, middle and near the end of each fuel cycle...."

The licensee states this is inconsistent with the actual specification which only requires the HTC to be measured near the beginning of the fuel cycle and again, near the end.

The licensee, therefore, proposes to correct this inconsistency by deleting the word " middle" from the Basis.

In support of this proposed change, the licensee states the revised wording'would also be consistent with that given in NUREG-0452, Revision 4 (Standard Technical Specifications for Westinghouse Pressurized Water Reactors).

We have reviewed the current wording of this specification and its related Basis. We have also reviewed the Standard Technical Specifications referenced by the licensee. Ba ed on this review, we conclude the proposed change would provide consistency with the present requirements of the technical specifications and with the Standard Technical Specifications..Accordingly, we find the i

proposed change acceptable.

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Basis - Specification 3/4'.2.1. _This specification sets forth the LCOs and Surveillance Requirements for Axial Flux Difference (AFD).

The licensee proposes to revise the associated. Basis section to correct a typographical error. The error involves the number =of detectors that must indicate the AFD is outside 'the t'arget ba'nd in order to initiate an alarm. The Basis presently states,an-alarm-message will be generated if 3 'of 4 or 2;of 3 OPERABLE excore channels are outside the target band;;while the specification' states the indicated AFD-shall be-considere'diout of its. target band when at least 2 of 4 or 2 of 3 OPERABLE channels"are indicating the AFD is outside the target band.

Based on'this' discrepancy'and the-fact a 2 of 4 requirement is more restrictive than a 3 of 4 requirement, we conclude the proposed e.

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change is a proper correction, does not reduce the present level of safety and is, therefore, acceptable, m.

Basis - Specification 3/4.4.3.2.

This specification sets forth Operability and Surveillance-Requirements for the pressurizer Power Operated Relief Valves (PORVs) and the associated Block Valves. The Basis for this specification presently states the PORVs have remotely operated block valves to provide a positive shutoff capability "should a relief valve become inoperable". The licensee states this wording may be misleading because there are situations, other than PORV inoperability, when it is necessary or desirable to close the Block Valve. These include situations.when the PORV has a minor leak, or when it is necessary to perform surveillance testing.

The licensee therefore proposes to revise the wording in the Basis to state the Block Valve can be closed "when necessary".

Based on our review, we agree the present wording in the Basis section is subject to mis-interpretation. We also agree there are conditions other than PORV inoperability when it is necessary or desirable to close the Block Valve. We further note Specification 3.4.3.2 imposes no limitations on closure of the Block Valves - only on valve operability. Accordingly, based on the foregoing considerations, we conclude the proposed change to this Baris Section is acceptable.

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Basis - Specification 3/4.4.4.

This specification addresses the Operability and Surveillance Requirements associated with the Pressurizer. One of the operability requirements is the pressurizer shall be OPERABLE with at least 150 kw of pressurizer heaters. The licensee proposes to revise the Basis section for this specification by adding a sentence stating that a minlmum of seven pressurizer heaters (23 kw each) are needed to meet this power requirement. We

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conclude this is an editorial revision and that it is acceptable.

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Basis - Specifications 3/4.5.2. and 3/4.5.3.1.

These specifications set forth the Operability and Surveillance Requirements for ECCS subsystems.

Specification 3.5.2 applies in Modes 1-3 and requires two independent ECCS subsystems to be OPERABLE in these Modes.

Specifi~ cation 3.5.3.1 applies in Mode 4 and requires only one ECCS subsystem to be OPERABLE. These' requirements are essentially identical to those stated in the Standard Technical Specifications for Westinghouse reactors. The licensee proposes to add to the Basis for these specifications an explanation of why only one ECCS subsystem is needed in Mode 4 (RCS temperature-< 350*F). The proposed explanation is virtually identical to that presented in the Standard Technical Specifications.

Inasmuch as'the proposed revision merely explains the Basis for existing requirements and does not reduce any present requirements or limitations, and because the proposed Basis is substantially identical with that presented in theLStandard Technical Specifications for Westinghouse reactors, we conclude the proposed revision is acceptable.

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Figure 6.2-2, Facility Organization. This Figure shows the administrative organizational structure at the facility and among other information, indicates those in the organization that are members of the Plant Review Board. The licensee states the revision of this Figure issued by Amendment 86 was in error because' it indicated Assistant Shift Supervisors were members of the Plant Review Board, and this is not consistent with Specification 6.5.1.2.

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The licensee. therefore' proposes to correct the error ~ by revising the 4

Figure as necessary to be consistent with Specification 6.5.1.2.

We-conclude this is an editorial change and is acceptable.

III. CONCLUSIONS Environmental Consideration i

This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.

The staff has determined that the amendment _ involves no significant increase in the amounts of any effluents that may be released offsite, j

and that there is no significant increase in individual or cumulative j

occupational radiation exposure., The Commission has'previously issued a proposed finding-that this amendment involves no significant hazards

' consideration and there has been no publicscomment_on.such finding.

i Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant t'o 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with issuance of this amendment.

Conclusion We have concluded, based on the considerations discussed above, that (1) 7.here is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such 4

activities will be conducted in compliance with the Commission's regulations, and the issuance of'the amendment will not be inimical to common defense and security or to the health and safety of the public.

j Date:

i Principal Contributor:

G. Zwetzig.

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