ML20209F446

Proposed Tech Specs Re Limiting Conditions for Operation of Protective Sys Instrument Setpoints & Allowable Values

ML20209F446
Person / Time
Site:	Fort Saint Vrain
Issue date:	06/21/1985
From:	PUBLIC SERVICE CO. OF COLORADO
To:
Shared Package
ML20209F429	List: ML20209F423 ML20209F440 ML20209F446
References
TAC-47416, NUDOCS 8507120469
Download: ML20209F446 (217)

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1 P-85214 ATTACHMENT 1 l Sumary of Proposed Changes

!O 8507120469 850621 PDR ADOCK 05000267 P PDR

r Attachment 1 P-85214 Summary of Proposed Changes to Fcrt St. Vrain Technical Specifications Section 2.1 " Definitions" Page 2-1 1. The following terms have been added to the " Definitions" Section:

2.la ACTION 2.lb ALLOWABLE VALUE 2.lc CHANNELS TO TRIP 2.1d MINIMUM CHANNELS OPERABLE 2.le OPERATIONAL MODE - MODE Page 2-la I. The following terms have been added to the " Definitions" Section:

2.lf TOTAL NO. OF CHANNELS 2.lg TRIP SETPOINT Page 2-9 I. Add Table 2.1-1 th Section 3.3 " Limiting Safety System Settings" Page 3.3-1 I. This page has been completely revised.

Page 3.3-2 I. An ALLOWABLE VALUE is added to each parameter.

II. Linear Channel-High (Neutron Flux) has the TRIP SETPOINT changed from less than or equal to 140% of RATED THERMAL POWER to varies as a function of Indicated Thermal Power with a note referencing curves to be approved by NFSC and submitted to the Commission each fuel cycle. The curves, High Neutron Flux Scram Detector Decalibration for Cycle 4, and High Neutron Flux Rod Withdrawal Prohibit Detector Decalibration curves for Cycle 4 present TRIP SETPOINTS and ALLOWABLE VALUES for all indicated power levels.

The curves and calibration instructions approved by NFSC for Cycle 4 are included as Attachment 5.

'~ III. Reheat Steam Temperature-High Scram has gi} a TRIP SETP0 INT change from less than l

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-Attachment 1 P-85214 or equal to 1075 degrees F to less than or equal to 1055 degrees F.

IV. Primary Coolant Pressure-Programed Low

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' Scram TRIP SETPOINT is changed from less than or equal to 50 psi below normal, programed with load, to less than or equal to 64.6 psi below normal, programed with Circulator Inlet Temperature.

V. A reference to Figure 3.3-1 is added for the Primary Coolant Pressure-Programed Low Scram ALLOWABLE VALUE to illustrate the programing of Primary Coolant Pressure to Circulator Inlet Temperature (an approximation for load).

Page 3.3-3 I. Primary Coolant Pressure-High Scram and pre-selected Loop Shutdown and Steam / Water Dump TRIP SETP0 INT has been changed from less than or equal to 53 psi above normal to less than or equal to 44 psi above normal, programed with Circulator Inlet Temperature.

II. Primary Coolant Moisture-High Scram, Loop Shutdown, and Steam / Water Dump has (m)

( a TRIP SETPOINT change from less than or equal to 67 degree F dewpoint to less than or equal to 60.5 degree F dewpoint.

III. An ALLOWABLE VALUE has been added for each parameter.

IV. The pressure relief valves and rupture discs TRIP SETPOINTS are unchanged.

l Page 3.3-4 I. This is a new page, however the TRIP SETP0 INT values are unchanged.

~Page 3.3-5 I. Note (a) is added to the Table of Notes of Sections 3.3 and 4.4.1 l Page 3.3-6 1. Figure 3.3-1 is a new figure being i added to better illustrate the

! programming of ALLOWABLE VALUES for l

Primary Coolant Pressure with l Circulator Inlet Temperature (an approximation for load).

Page 3.3-7 I. This is a new page to introduce the O b ses for 'sss 3 3-

r Attachment 1 P-85214 Page 3.3-8 I. Basis is changed to bases.

II. Trip Setting is changed to TRIP Q SETPOINT.

III. The title is changed from High Neutron Flux to Linear Channel - High (Neutron Flux).

IV. "However, near-RATED THERMAL POWER" is replaced with "During normal power operation."

V. RATED THERMAL POWER has been capitalized.

VI. The discussion of the 140% analysis value for High Neutron Flux has been revised to address the changing of the TRIP SETPOINT to account for neutron flux detector decalibration.

4 VII. The title of High Reheat Steam Temperature is changed slightly.

Page 3.3-9 I. In the basis for Primary Coolant Pressure-Programmed Low, Trip Setting is changed to TRIP SETP0INTS.

O II. 1ne titie of High erimery Cooient Pressure is changed slightly.

III. The 12-second time period is qualified by adding the phrase " assuming rated power and flow conditions."

IV. The language is changed from "is limited to 705 psia" to "does not exceed 705 psia" and reference to FSAR Table 14.5-1 is eliminated.

V. A discussion of backup protection is removed and the trip setting of less than or equal to 53 psia is replaced with a TRIP SETPOINT of less than or equal to 44 psi above normal operating pressure. The backup protection discussion is moved to Primary Coolant Moisture-High.

VI. The phrase "for a time interval of approximately 75 seconds in the case of the maximum, leak rate" is removed.

Q VII. A reference to FSAR Table 14.5-1 is removed. ,

r Attachment 1 P-85214 VIII. A reference to Figure 3.3-1, an addition to the Technical Specification, is added.

O IX. The upper trip setting limit at 775 psia at a programmed Circulator Inlet Temperature of 700 degree F with reference to FSAR Figure 7.1-14 has been removed.

X. The title High Moisture in the Primary Coolant is changed slightly.

XI. Tha TRIP SETPOINT is changed from less thar or equal to 500 ppmv to .60.5 degrees F dewpoint. The second reference to 500 ppmv is deleted.

Page 3.3-10 I. A discussion of backup protective action by Primary Coolant Pressure-High Scram is added.

II. A minor change in language and reference to LC0 4.2.7 is added.

Section 4.0 Page 4.0-2 I. This page is added to be consistent p with the format of the Westinghouse V Standard Technical Specifications and to address the use of ACTION Statements for LC0 4.4.1 only.

Section 4.4 Page 4.4-1 1. This page is completely revised to be consistent with the format of the Westinghouse Standard Technical Specifications.

Page 4.4-2 1. Old page 4.4-2 is eliminated and replaced with Table 4.4-1 (Part 1).

II. "(Part 1)" is added to the table i description to differentiate between the Setpoints portion of the table and the operating requirements portion of the table.

III. " Trip Setting" is changed to " TRIP SETP0 INT" and an " ALLOWABLE VALUE" column is added and other columns as modified are transferred to Part 2 of the table.

O IV. Change items la and Ib to " Manual Scram" from " Manual."

r Attachment 1 P-85214 V. Change STARTUP Channel High TRIP SETPOINT from less than or equal to 1.0E+05 cps less than or equal to O 8.3E+04 cps.

VI. Add item 2b " Wide Range Channel Rate of Change-High" with TRIP SETP0 INT and ALLOWABLE VALUE less than or equal to 4.5 decades.per minute (dpm).

VII. Change TRIP SETPOINTS for Linear Channel-High (items 3a and 3b) and add a note referencing curves to be approved by NFSC and submitted to the Commission. The curves, High Neutron Flux Scram Detector Decalibration curves for Cycle 4, will present TRIP SETPOINTS and ALLOWABLE VALUES for all indicated power levels.

VIII. The TRIP SETPOINT for Primary Coolant Moisture-High Level Monitor is changed to less than or equal to 60.5 degree F dewpoint from 67 degree F dewpoint.

IX. The TRIP SETPOINT for Primary Coolant Moisture-Loop Monitor is changed to less than or equal to 20.4 degree F g s. dewpoint from 27 degree F dewpoint.

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X. The TRIP SETP0 INT for Reheat Steam Temperature-High is changed from less than or equal to 1075 degree F to less than or equal to 1055 degree F.

Page 4.4-2a I. The TRIP SETPOINT for Primary Coolant Pressure-Programmed Low is changed from less than or equal to 50 psig below nomal, load programmed to less than or equal to 64.6 psi below normal, programmed with Circulator Inlet Temperature.

II. The TRIP SETPOINT is changed for Primary Coolant Pressure-Programmed High from less than or equal to 7.5%

above normal rated, load programmed, to less than or equal to 44 psi above normal, programmed with Circulator Inlet Temperature. An upper TRIP SETP0 INT of 744 psia is also added.

. III. The Hot Reheat Header Pressure-Low TRIP SETP0 INT is changed from greater than or equal to 35 psig to greater than or Q equal to 44 psig.

e-Attachment 1 P-85214 IV. The Main Steam Pressure-Low TRIP SETPOINT is changed from greater than or equal to 1500 psig to greater than (v] or equal to 1529 psig.

V. The Plant Electrical System-Loss TRIP SETPOINT is changed from loss of voltage on both Bus 1A and 1C for more than 35 seconds to greater than or equal to 278 volts indicated by two of three relays on two of three 480V busses.

VI. Two Loop Trouble TRIP SETPOINT is unchanged as not applicable.

VII. The TRIP SETPOINT for High Reactor Building Temperature (Pipe Lavity) is changed from less than or equal to 325 degree F to less than or equal to 161 degree F.

Page 4.4-2b I. Columns used in existing FSV Technical Specifications labeled " Minimum Operable Channels", " Minimum Degree of Redundancy", and " Permissible Bypass Conditions" are changed to " TOTAL NO.

OF CHANNELS", " CHANNELS TO TRIP",

CHANNELS OPERABLE",

" MINIMUM

'O " APPLICABLE MODES", and " ACTION".

II. Add new information corresponding to the new column headings. Only MINIMUM CHANNELS OPERABLE column remains unchanged. The S/U Applicable MODE has been removed from Functional Unit 4 and 5 as explained in Attachment 3 pages 24 and 25.

Page 4.4-2c I. The column headings are changed as they are on the previous page. The S/U Applicable MODE has been removed from Functional Unit 7 as explained in Attachment 3 pages 25 and 26.

Page 4.4-2d I. The notes for the tables have been divided such that each table has its notes immediately following it rather than one set of notes for all tables.

II. Note (a) is a new note.

III. The wording of note (b) is changed slightly from previous note (a).

IV. Note (c) is a clarification of previous note (d).

r Attachment 1 P-85214 V. Note (d) is a new note to indicate the limited applicability of the scram controls within the refueling MODE.

V VI. Note (e) is a simplified version of previous note (c).

VII. Note (f) is the same as previous note (t) with some of the symbols replaced with words.

Page 4.4-2e I. Note (g) is identical to previous note (h).

II. Note (h) is identical to previous note (b).

III. Note (i) is similar to previous note (k) with the second "0PERABLE" changed to " Operating."

IV. Note (j) is a new note.

V. Note (k) is a new note.

Page 4.4-2f and I. The ACTION statements were included in 4.4-2g the FSV Technical Specifications as notes or in ~the introduction at the beginning of Section 4.4. In O accordance with the Standard Technical Specification Format, ACTIONS are now defined as ACTION STATEMENTS for each table and numbered for use in that table.

Page 4.4-3 I. The Steam Pipe Rupture Under PCRV, Loop 1/ Loop 2 TRIP SETP0 INT is changed from less than or equal to 9 VDC to less than or equal to 8.68 VDC.

II. Allowable Values are added for each Functional Unit as in Table 4.4-1.

III. The Steam Pipe Rupture, North / South Pipe Cavity, Loop 1/ Loop 2 TRIP SETPOINTS are changed from less than or equal to 9 VDC to less than or equal to 8.68 VDC.

IV. The High Pressure, Pipe Cavity TRIP SETPOINT is changed from less than or equal to 2.5" H2O to less than or equal to 1.3" H20.

p V. The High Temperature, Pipe Cavity, TRIP v SETP0 INT is changed from less than or

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Attachment 1 P-85214 equal to 130 degree F to less than or equal to 125 degree F.

VI. The High Pressure, Under PCRV, TRIP SETP0 INT is changed from less than or equal to 2.5" H20 to less than or equal to 1.3" H20.

VII. The High Temperature, Under PCRV, TRIP SETPOINT is changed from less than or equal to 130 degree F to less than or equal to 125 degree F.

VIII. The logic circuits are changed to read "Not Applicable" for the TRIP SETP0INTS and ALLOWABLE VALUES. The logic circuits include Loop 1 Shutdown Logic, Loop 2 Shutdown Logic and Circulator 1A and 1B Shutdown Logic.

Page 4.4-3a I. The Circulator 1C and 1D Shutdown Logic is labeled "Not Applicable" for both the TRIP SETP0 INT and ALLOWABLE VALUE.

II. The Steam Generator Penetration Overpressure, Loop 1/ Loop 2 TRIP SETPOINT is changed from less than or equal to 810 psig to less than or equal to 788 psig.

O III. The High Reheat Header Activity, Loop 1/ Loop 2, TRIP SETPOINTS are changed from less than or equal to 5 mr/hr to less than or equal to 3.2 mr/hr.

IV. The Low Superheat Header Temperature, Loop 1 and Loop 2, TRIP SETP0 INT is changed from less than or equal to 800 degree F to less than or equal to 798 degree F.

V. The High Differential Temperature Between Loop 1 and Loop 2 TRIP SETPOINT is changed from less than or equal to 50 degree F to less than or equal to 44.8 degree F.

VI. Item 8, Primary Coolant Moisture, is added with reference to Table 4.4-1 for TRIP SETPOINT and ALLOWABLE VALUE.

Pages 4.4-3b, 4.4-3c I. The column headings on these pages are and 4.4-3d changed to match those used in the g Westinghouse Standard Technical V Specifications.

Attachment 1 P-85214 II. Only the MINIMUM CHANNELS OPERABLE column remains the same as the original FSV Technical Specifications.

gV III. Each of the column headings is defined in Section 2.0.

Page 4.4-3e I. Note (1) is a rewrite of former note (j) to clarify the note and replace number designations with words.

II. Note (m) is a rewrite of former note (p) to clarify the note and replace number designations with words.

III. Note (n) is similar to former note (s).

IV. Note (o) is a new note describing the Applicable MODES for the Circulator Loop Shutdown Logic.

V. Notes (p) and (q) are intentionally left blank for future use.

Page 4.4-3f I. This page is completely new and defines the new ACTION STATEMENTS used on Table 4.4-2 (Part 2).

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p Pages 4.4-4, I. Table 4.4-3 (Part 1) has been revised O 4.4-4a and to only include the TRIP SETPOINT and 4.4-4b ALLOWABLE VALUE for each functional unit.

II. The functional units have been reorganized to group those items associated only with circulator drive on water and only with circulator drive on steam, then followed by common

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items.

III. The Manual Trip (Water) functional unit item has been added.

IV. Functional unit items for Loop 1 and Loop 2 " Programmed Feedwater Flow-Low" have been'added for 2 circulators in a loop and 1 circulator in a loop.

V. Manual Trip (Steam) has been changed to read "Not Applicable" for both TRIP SETP0 INT and ALLOWABLE VALUE.

VI. Circulator Speed-High (Steam) has a revised TRIP SETP0 INT from < 11,000 rpm to < 11,495 rpm. An ANALYTIS VALUE of

(] 11,7U0 rpm was used as submitted in P-84137.

r Attachment 1 P-85214 VII. Circulator Drain Malfunction has a revisod TRIP SETPOINT from > 5 psid to

-> 8 ,)sid.

. ,7 U VIII. Circulator Speed-High (Water) has a reiised TRIP SETP0 INT from < 8,800 rpm to 1 8,589 rpm.

IX. Circulator Speed-Low has a revised TRIP St.TPOINT from 1910 rpm below normal to 1850 rpm below nonnal.

X. Loop 1 (Lep 2), Fixed Feedwater Flow-Low has a revised TRIP SETP0 INT from 20% of normal Full Load to > 15.4% of-normal full load or _>_ 177,500 lb/hr.

XI. A TRIP SETP0 INT of < 211,000 lb/hr (18.3%) below normal as progranined by circulator speed for either 2 circulators per loop or 1 circulator per loop was established for Programmed Feedwater Flow-Low.

XII. Circulator Seal Malfunction has been.

revised to have a low TRIP SETP0 INT of

> -6" H2O instead of > -10" H20 with a liew ALLOWABLE VALUE of-'_) -6" H20.

g V XIII. A separate functional unit listing has been designated for Circulator Seal Malfunction-High with a TRIP SETPOINT of < 75.5" H2O instead of < 80" H20.

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XIV. Loss of Circulator Bearing Water TRIP SETPOINT has been revised from > 475 -

osid to > 459 psid.

XV. Circulator Penetration Trouble TRIP SETPOINT has been revised from ~~ < 810 psig to i 786 psig.

Pages 4.4-4c and I. These pages are completely new to 4.4-4d include Table 4.4-3 (Part 2) that incluies instrument operating requirements.

II. The TOTAL N0. OF CHANNELS column, CHANNELS TO TRIP column, and the ACTION column are new to these Technical Specifications.

III. The APPLICABLE MODES column replaces the Permissable Bypass Conditions column.

n Attachment 1 P-85214 IV. The MINIMUM CHANNELS OPERABLE column replaces the Minimum Operable Channels column but includes the same C information with the addition of similar information for the new functional units.

Pages 4.4-4e and I. These are new Figures 4.4-la and 4.4-4f 4.4-lb with curves designed to provide information on the programing of the ALLOWABLE VALUES for changes in Circulator Speed and Feedwater Flow rate.

Page 4.4-49 I. Note (r) is a new note required to better define the applicable MODES for the circulator protection functional units.

II. Note (s) is a new note to clarify the applicable MODES for Circulator Drain Malfunction protection.

III. Note (t) is a new note to clarify the applicable MODES and MINIMUM CHANNELS OPERABLE for Circulator Seal Malfunction.

IV. Notes (u), (v) and (w) have been O intentionally left open for future use.

Page 4.4-4h I. This is a new page to define ACTION STATEMENTS that are new to this revision.

Page 4.4-5 and I. Table 4.4-4 (Part 1) replaces the 4.4-Sa Trip Settings with TRIP SETP0 INT and adds an ALLOWABLE VALUE for each j functional unit. The other information on instrument operating requirements is l ~

addressed in Table 4.4-4 (Part 2).

II. The STARTUP Channel-Low Count Rate (Channels 1 and 2) TRIP SETP0 INT has been changed from > 2.5 cps to > 4.2 cps.

t III. The Linear Channel-5% RWP (Channels 3, l

' 4 and 5 and Channels 6, 7 and 8) TRIP i SETP0INTS have been changed from > 5%

l of RATED THERMAL POWER to i 5% of RXTED l THERMAL POWER.

l IV. The Linear Channel-30% RWP (Channels 3, 4 and 5 and Channels 6, 7 and 8) TRIP

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k SETP0INTS are unchanged at i30%of RATED THERMAL POWER.

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r Attachment 1 P-85214 V. New functional units have been added to the Rod Withdrawal Prohibits, including

1) Wide Range Channel Rate of Change-High, 2) STARTOP Channels Rate of O'V Change-High, and 3) Linear Channel-High power RWP's for Channels 3, 4 and 5 and for Channels 6, 7 and 8.

VI. . Add Item 6 Multiple Rod Pair Withdrawal.

Pages 4.4-5b and I. Table 4.4-4 (Part 2) is a new 4.4-Sc presentation of Instrument Operating Requirements. The column for TOTAL N0.

OF CHANNELS, CHANNELS TO TRIP, and ACTION are completely new and replace the Minimum Degree of Redundancy column.

II. MINIM 0M OPERABLE' CHANNELS is replaced with MINIMUM CHANNELS OPdRABLE with no change of data. The APPLICABLE MODES column replaces the Pennissible Gypass Conditions column.

III. As in Table 4.4-4 (Part 1) new functional units have been added including 1) Wide Range Channel Rate of 7 Change-High (Channels 3, 4 and 5), 2) th, STARTUP Channels Rate of Change-High (Channels 1 and 2),~3) Linear Channel-High Power RWP (Channels 3, 4 and 5 and Channels o, 7 and 8), and 4) Multiple Rod Pair Withdrawal.

Pages 4.4-5d I. Note (x) is a version of note (m) revised to improve clarity.

II. Note (y) is a version of note (n) revised to improved clarity.

III. Note (z) is a new note and is similar to note (a) on Table 4.4-1 and note (a) on Table 3.3-1.

IV. Note (aa) is a new note defining limitations on the applicability of the RWP in the refueling MODE.

V. Note (bb) is a new note defining limitations on the applicability of the RWP in the refueling MODE.

VI. Note (cc) has been left open C intentionally for future use.

I Attachment 1 P-85214 Page 4.4-Se 1. The ACTION STATEMENTS are new with this amendment to the Technical Specifications.

O Page 4.4-6 1. This page has not been included in any previously approved amendment to these Technical Specifications.

II. Plant Electrical System-Loss (Scram) is the same functional unit that appears in Table 4.4-1.

Pages 4.4-6a, I. These three pages are all new 4.4-6b and along with Table 4.4-5.

4.4-6c II. Note (dd) is a new note.

III. Note (ee) is the same as note (c) of Table 4.4-1. Both notes are an improved version of note (d).

IV. Notes (ff)and(gg)havebeenleftopen intentionally for future use.

V. The ACTION STATEMENTS are new with this Amendment to the Technical Specifications.

O Page 4.4-7 I. Page 4.4-7 is a new page to introduce the Bases and separate them from the Technical Specifications.

Pages 4.4-8 I. These pages have been revised to through include e, discussion of the Basis for 4.13e each protective ACTION in Tables 4.4-1 through 4.4-5, time allowed for actions and Drift.

II. The Bases have been extensively revised to improve clarity.

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i P-85214 ATTACHMENT 2 Proposed Changes O

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m Fort St. Vrain #1 Technical Spacifications

- Amendment No.

Page 2-1 2.0 Definitions The following frequently used terms are defined to provide a uniform basis for interpretation of these Technical Specifications.

l 2.1 THREE ROOM CONTROL COMPLEX That area of the Turbine building which includes the Control Room, the Auxiliary Electric Room and the 480 Volt Switchgear Room.

l 2.la ACTION ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions.

l 2.lb ALLOWABLE VALUE The ALLOWABLE VALUE shall be the least conservative acceptable "as found" value for a TRIP SETPOINT.

l 2.lc CHANNELS TO TRIP

, The CHANNELS TO TRIP shall be the number of channels that are required to be actuated in order to initiate a O

Protective actioa.

l 2.1d MINIMUM CHANNELS OPERABLE The MINIMUM CHANNELS OPERABLE shall be that number of channels required to be OPERABLE in the applicable MODES. Failure to restore the number of channels required to the OPERABLE status, to meet the stated MINIMUM CHANNELS OPERABLE, shall invoke an ACTION usually requiring reactor shutdown, loop shutdown, circulator shutdown, or Rod l

Withdrawal Prohibit activated after a stated grace period.

l 2.le OPERATIONAL MODE - MODE These definitions apply only to LSSS 3.3, LCO 4.4.1 and SR 5.4.1, and supercede definitions 2.5, 2.10, 2.14 and 2.16 for those sections.

DPERATIONAL MODE (i.e. MODE) shall correspond to any one inclusive combination of reactor MODE switch setting, l

Interlock Sequence Switch Setting, and Power Level, specified in Table 2.1-1.

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r i Fort St. Vrain #1 Technical Specifications Amendment No. 1 Page 2-la l 2.lf TOTAL NO. OF CHANNELS

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The TOTAL NO. OF CHANNELS shall designate the sum of channels installed to provide a trip signal to a protective device. j r

l 2.lg TRIP SETPOINT The TRIP SETPOINT is the least conservative "as left" value (as indicated) on a protective device to prevent a measured quantity from exceeding the ALLOWABLE VALUE.

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Fort St. Vrain #1 Technical Specifications Amendment No.

Page 2-9 O 2.0. Definitions l TABLE 2.1-1 l OPERATIONAL MODES The modes defined below apply only to LSSS 3.3, LC0 4.4.1 and SR 5.4.1, and supercede definitions 2.5 Low Power Operation, 2.10 Power Operation , 2.14 Reactor Shutdown, and 2.16 Refueling Shutdown for those sections only.

INTERLOCK SEQUENCE REACTOR MODE  % INDICATED MODE (1) SWITCH SETTING SWITCH SETTING THERMAL POWER (2)

(P) POWER Run >30%

l POWER LOW POWER Run 130%and>5%

lLOWPOWER(L)

STARTUP Run 15%

lSTARTUP (S/U) l SHUTDOWN (S/D) (3) Off 0%

(3) (4) Fuel Loading 0%

Q l REFUELING (R) 1 (1) All three conditions as defined by the Interlock Sequence Switch, Reactor Mode Switch and % Indicated Thermal Power define the Applicable MODE.

l (2) Excluding Decay Heat l.(3) ISS may be in any position in SHUTDOWN or REFUELING MODES i (4) Refueling includes reactor internal maintenance, see Specificaton LC0 4.5.2'.

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r Fort St. Vrain #1 Technical Specifications Amendment No.

Page 3.3-1 S^" S'S'5" S'" '"SS OI- ""

l OBJECTIVE:

The' Plant Protective System (PPS) instrumentation and PCRV Pressurization TRIP SETPOINTS shall be set consistent with the TRIP SETPOINT values shown in Table 3.3-1.

l APPLICABILITY: As shown for FUNCTIONAL UNIT in Table 4.4-1.

l ACTION:

With a Limiting Safety System Setting less conservative than the value shown in the ALLOWABLE VALUE column of Table 3.3-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Limiting Condition for Operation 4.4.1, PPS, or 4.2.7, PCRV Pressurization.

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Fort St. Vrain #1 Technical Specifications Amendment No.

Page 3.3-2 Specification LSSS 3.3

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l Table 3.3-1 l LIMITING SAFETY SYSTEM SETTINGS TRIP ALLOWABLE PARAMETER FUNCTION SETPOINT VALUE

1. Reactor Core Limiting Safety System Settings a) Linear Scram Varies as a Varies as a Channel-High Function of Function of (Neutron Indicated Indicated Flux) Thermal Thermal Power (a) Power (a)

Scram < 1055 < 1061 b) Reheat Fegree F Steam Fegree F Temperature-High l '

c) Primary Scram < 64.6 psi < 67 psi Coolant Felow normal, Felow normal, programmed

, O. Pressure-Programmed programmed with Circu- with Circu-Low lator Inlet lator Inlet Temperature. Temperature Upper TRIP per Figure SETPOINT of 3.3-1. Upper

> 635.4 psia. limit to produce trip at > 633 psia.

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Fort St. Vrain #1 Technical Specifications Amendment No.

Page 3.3-3 Oi Specificatioa LSSS 3.3 l Table 3.3-1 (Continued) l LIMITING SAFETY SYSTEM SETTINGS TRIP ALLOWABLE PARAMETER FUNCTION SETPOINT VALUE

2. Reactor Vessel Pressure Limiting Safety System Settings a) Primary Scram and < 44 psi < 47 psi Coolant Preselected libove normal, libove normal.

Pressure- Loop Shutdown programmed programmed Progranned and Steam / with Circu- with Circu-High Water Dump lator Inlet lator Inlet Temperature. Temperature Upper TRIP per Figure SETPOINT of 3.3-1. Upper

' < 744 psia, limit to Eower TRIP produce trip SETPOINT of at < 747 i

< 536 psia. psili. Lower

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• limit to produce trip at < 539 psia.

b) Primary Scram, Loop < 60.5 < 60.5 Coolant Shutdown, Hegree F Hegree F Moisture- and Steam / dewpoint dewpoint High Water Dump temperature temperature c)PCRV Pressure Pressure: Relief Rupture Disc 812 psig plus 820 psig (Low Set or minus 8 Safety Valve) psi O

Fort St. Vrain #1 Technical Specifications Amendment No.

Page 3.3-4 Oi Specification tSSS 3.3 l

Table 3.3-1 (Continued) l LIMITING SAFETY SYSTEM SETTINGS TRIP ALLOWABLE PARAMETER FUNCTION SETPOINT VALUE Low Set Safety 796 psig plus 804 psig Valve or minus 8 psi Rupture Disc 832 psig plus 840 psig (High Set Safety or minus 8 psi Valve)

High Set Safety 812 psig plus 820 psig Valve or minus 8 psi d) Helium Pressure Circulator Relief Penetration Interspace Pressure:

825 ,sig pius 842 psig O auPture oisc (2Per or minus 17 Penetration) psi Safety Valve 805 psig plus 829 psig (2 Per or minus 24 Penetration) psi e) Steam Pressure Generator Relief Penetration Interspace Pressure:

Rupture Disc 825 psig plus 842 psig (2 For Each or minus 17 Steam Generator) psi Safety Valve 475 psig plus 489 psig (2 For Each or minus 14 Steam Generator) psi O

T Fort St. Vrain #1 Technical Specifications Amendr. tent No.

Page 3.3-5 Table 3.3-1 (CONTINUED)

Ql TABLE NOTATION l

(a) Curves ssecifying the Linear Channel High Neutron Flux TRIP SETPOINT 'imits and ALLOWABLE VALUE as a function of indicated power level, which account fer neutron detector decalibration, shall be established for each fuel cycle. The neutron detector decalibration curves and instructions shall be approved by the NFSC prior to each fuel cycle. The detector decalibration curves approved by the NFSC shall be forwarded within 30 days of approval to the Regional Administrator, Region IV, The Comission. See Tables 4.4-1 an(. 4.4-4 for related limits.

s O

O

Fort St. Vrain Technical Specifications Arr.endment No.

Page 3.3-6 O

r-Y I

I I

L K

s00 ~ ALLOWABLE t- t

HIGH PRIMARY h= === H~

Q COOLANT :i _. . _J PRESSURE 4- 747 PSIA ' ~.4_-_

55 L->

PRESS.=.6338 :i-S g (TEMP)+ 276.7 iF =W7hDSIA -~

a  ! i:  :

% ~ 742 F =

1. NORMNC ~E_3 W

'\i'--'f~=&}- - OPERATING ZZ .

[ " "N P PRESSURE E n mt:---

. /- i Y /y ._633 ~"

PSIA t:===

y~-

/  : j ALLOWABLE m

8 M '

NX u

eaa '

d - E$*'"T=f~' LOW c*gu,i N

, PRIMAR g __._[_. /_ _/ -- - L--. - +- .t~=--

_..___{=---

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4 5 539 PSIA __.__f

~g= LOW PRESSURE

{ -Q SCRAM BYPASSED ____ .l_:_ _._..,

m -

=l~- + .

WHEN NOT IN THE ~ ~ ~ ' -- : =i-

" POWER MODE EEdEEE5;.

i  :

s00 I i . I l i I s00 600 700 s00 I 300 400 CIRCULATOR INLET TEMPERATURE (oF)

FIGURE 3.3-1 PRIMARY COOLANT PRESSURE vs. CIRCULATOR INLET TEMPERATURE l

ALLOWABLE OPERATION O

Fort St. Vrain #1 Technical Specifications Amendment No.

4 Page 3.3-7 Oi note The Bases contained in succeeding pages sumarize the reasons for the Specifications in Section 3.3, but in accordance with 10CFR50.36 are i not part of these Technical Specifications.

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Fort St. Vrain #1 Technice.1 Specifications

- Amendment No.

4 ,

.* Page 3.3-8 ,!

lBasesforSpecificItionLSSS3.3. ,

f 4 , .* % ,9 i ., N -

Limits. established in 15pecificttionbc. SL 3.1 and SL 3.2 (f$6fety

&' s. : e safeguard the fuel particle iategrity. and 4he readtor coolant system h tN barriers. Protective devices included in LSSS 3.3 have been provided Yi in the plaint design to ensure that automatf ? corrective action is j jtaken when Yequired to prevent ,the Limiting Safety System Settings from being -ekceede( during, normal operation,' 'orL during operational transients re ulting from possible operator arrors, orias a result of

,t , eqtipment malfunction. This Jspecificatio$ establishes the TRIP SETPOINTS ard ALLOWABLE VALUb for thc'se . automatic '

protective devices. . t,j-y ,b.

f0peation with 'a trip set lessiconservative than its TRIP SETPOINT

't)but within its specified ALLOWABLE YALUE is acceptable ~on the basis

!- - I sthat the' , difference between each TRIP SETPDINT and the ALLOWABLE l

s j VALUE % eoual to or less than the drift allowance for all trips includhg those trips assumed in the saihty analyses.

+ c 4 i

3 ,

g'-Where the drift is. only measured by a CHAMEL CALIBRATION no margin is showi befween the TRIP SETPOINT and ALLOVABLE VALUE.

y^ , ,

Nigh-(deutron fivd{ , Th s[ W ll Linear Cha'n ul

- , s af .7 l> .

( N{ ' l TMt neutron,f)sxf TRIP SETPOINTS ere established to protect the fuel Mrticla' integrity during rapid tuerpower itransknts. The power r h[C.

i range' nuclear dannels respond,t(changes b neutron flux. During l 1

h , normal power *cperation, the channels.are -cal,ibrate'd using a plant heat balance so that the neutirn flax that is' sensed is read out as .

i Ir lpercentofRATEDTHERMALPOWEsJ.Jfar/slowmaneuvers,thosewherecore

thermal powe G surface heat lui?, and the power transferred to the h helium follow.the neutren flux,(the power range nuclear channels will

[NO)d indicate reactor. thermal'. power. For fast transients, the neutron i' ~ * !

flux change will lea 6 the crar.ge in power transferred from the core

~

to the helium due to the effect of the fuel, moderator and reflector thermal time constants. Therefore, when the neutron flux increases v l to the scram TRIP SETT0 INT rapidly, the percent increase in heat flux

!- and power transferred to the helium will'be less than the percent increase in neutron iflux. . TRIP SE100INTS that assure a reactor

T}- scrar at no grdater thi.n'140F of RATED THIRMAL POWER are sufficient 2 for the 'plari ' because the nentive temperature coefficient of reactivity and lary her.t capaci@of the reactor limit the transient y,$

increases in fuel add' helium temperatures to acceptable values.

~'

Control rod shim bar:k mcvement cae result in decalibration of the ex-core flux detect. ors. / To' account ib'r this potentialsdecalibration, K the actual TRIP'SETPOINT is' administratively set less\ than 140% of THERMAt POWER based ,,upon indicated power. These i3\ RATED administratively set fitx TRIP SETPOINTS assure the scram will occur at or less than 140% of RAW D THERMAL POWER for those postulated reactivity accidents evaluated- in2 FSAR-Section 14.2. . 4 <

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r Fort St. Vrain #1 Technical Specifications Amendment No.

lReheatSteamTemperature-High

-High reheat steam temperature indicates either an increase in thermal power generation without an appropriate increase in helium cooling flow rate or a decrease in steam flow rate. Reheat steam temperature in lieu of reactor core outlet helium temperature is used because of the difficulty in measuring gross helium temperature for protective system purposes. The design of the steam generator is such that changes in . hot helium temperature due to a power increase first affect the reheat steam temperature thus allowing the latter to serve as an index of the. helium temperature. A reheat steam temperature scram is provided to prevent excessive ratio of power-to-helium flow due to a power increase or steam flow imbalance. (Section 14.2 of

~

the FSAR.)

Primary Coolant Pressure - Programed Low l The low primary coolant pressure TRIP SETPOINT has been established to maintain the fuel particle coating integrity due to loss of primary coolant as the result of a coolant leak.

l Primary Coolant Pressure - Programmed High The major potential source of primary coolant pressure increase above Q the. normal operating range is due to water and/or steam leakage by means of a defective evaporator-economizer-superheater subheader or For a double ended offset tube rupture, the rate of water and tube.

steam leakage will not exceed 35 lbs/sec initially, resulting -in a maximum rate of primary coolant pressure rise of approximately 1 psi per second. The normal plant protection system action upon detection of moisture is reactor scram, loop shutdown, and steam / water dump (FSAR Section7.1.2.5), occurring after approximately 12 seconds, assuming rated power and flow conditions. In this situation, the peak PCRV pressure at 100% reactor power does not exceed 705 psia.

The TRIP SETP0 INT of <44 psi above the normal operating pressure between 25% and 10U% of rated power is selected: (1) to prevent false scrams due to normal plant transients, and (2) to allow adequate time for the normal protective action (high moisture) to terminate the accident while limiting the resulting peak PCRV pressure in the unlikely event that the normal protective action were inoperative. In this case, reactor pressure would continue to rise to the high pressure TRIP SETPOINT. The resulting peak PCRV pressure would be less than the PCRV reference' design pressure. The high pressure TRIP SETPOINT is programmed as a function of load, using helium circulator inlet temperature as the measured variable jindicative of load, as shown in Figure'3.3 -1. The PCRV safety valves provide the ultimate protection against primary coolant system pressure exceeding the PCRV reference design pressure of 845 psig.

lPrimaryCoolantMoisture-High The high moisture TRIP SETPOINT corresponding to 60.5 degree F dewpoint was established, considering the moisture monitor characteristics and the necessity to minimize water leakage to the

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

Fort St. Vrain #1 Technical Specifications Amendment No.

. . Page 3.3-10 t .lreactorsystem. A trip would be reached after several hours of full power _ operation with a minimum water / steam inleakage rate in excess lofabout20lbs/hr. Below that leakage rate, the TRIP SETPOINT would never be reached, but the indicating instruments would show an abnormal condition. For maximum design leakage rates,- the system behavior. is as discussed in the preceding section on Primary Coolant Pressure-Programmed High. Backup protective action- is provided by i- the high primary coolant pressure scram, loop shutdown, and dump of a pre-selected -and remaining loop steam depressurization-Programmed High. loop (FSAR Sections 7.1.2.3 and 7.1.2.4).

PCRV Pressure

.If - the. pressure in the PCRV were to rise significantly above the normal operating pressure, the low-set rupture disc would rupture within the range of 804 psig (-1%), to 820 psig (+1%). The low set

safety ' valve, set at 796 psig + 1%, would be wide open and flowing full capacity at or above 820 psig (3% accumulation). If the pressure still continued to rise, the high-set rupture disc would rupture between 824 psig and 840 psig. The high-set safety valve, set at 812 psig + 1%, would be flowing full capacity above 836 psig l (3% ' accumulatioii). As the pressure decreased, the high-set safety valve would close at a pressure of approximately 690 psig and the low-set safety valve at approximately 677 psig; the corresponding

. ~Q: l primary set safetysystem valve closed. pressure would (FSAR be6.8.3).

Section approximately See also LC0 737 psig when the low-4.2.7.

Helium Circulator Penetration Interspace Pressure

! The penetration interspaces are protected against pressures exceeding PCRV reference pressure. The safety valves are set at 805 psig and rupture discs are set at 825 psig (nominal). A redundant safety

' valve and rupture disc are provided. The rupture discs would burst in the pressure range of 809 psig (-2%) to 842 psig (+2%). The to 829 psi safety valves would open in the range of 781 psig (-3%)(+3%) and would reliev .

l-

' The safety _ valves would reseat -at about 725 psig. The safety valve and rupture disc relieving pressures were specified so as to comply with the ASME Boiler and Pressure Vessel Code,Section III, Class B, l Nuclear Vessels, for over pressure protection. See also LC0 4.2.7.

Steam Generator Penetration Interspace Pressure The six steam generator penetration interspaces in each loop are provided with common upstream rupture discs and safety valves to protect a pressures exceeding PCRV reference design pressure (845 psig) gainst. A redundant safety valve and rupture disc are .provided.

The rupture discs would burst in the pressure range of 809 psig (-2%)

to 842 psig (+2%), with a nominal setting of 825 psig. The safety valves are each set at 475 psig which allows for a pressure drop in

.the' inlet lines of 370 psi when relieving at valve capacity. See also LC0 4.2.7.

l

[

Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.0-2 4.'0 LIMITING CONDITIONS FOR OPERATION (Applies only to LC0 4.4.1) 4.0.1 The Limiting Conditions for Operation specified in this section define the lowest functional capability or performance levels necessary to assure safe. operation of the. facility. These Limiting Conditions for Operation provide for operation with sufficient redundancy so that further, but limited, degradation of equipment capability or performance, or the occurrence of a postulated incident will not prevent a safe reactor shutdown.

4.0.2 These Liniiting Conditions for Operation do not replace plant operating procedures. Plant operating procedures establish plant operating conditions with at least the capability and performance specified in these Limiting Conditions for Operation.

4.0.3 Compliance with the Limiting Conditions for Operation contained in the succeeding Specifications is required during the OPERATIONAL MODES cr other conditions specified therein, except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

O 4.0.4 Noncompliance with a Specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is. restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

4.0.5 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, action shall be initiated to place the unit in a MODE in which the Specification does not apply by placing it, as applicable, in at least STARTUP within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in at least SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

This Specification is not applicable in SHUTDOWN or REFUELING except where otherwise noted.

4.0.6 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions for the Limiting Conditions for Operation are met without reliance -on provisions contained in the ACTION requirements. This provision shall not prevent passage through or to OPERATIONAL MODES of a lower power level. as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual Specifications. -

Fort St. Vrain #1 Technical Specifications Amendment No.

l4.4INSTRUMENTATIONANDCONTROLSYSTEMS l PLANT PROTECTIVE SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 4.4.1 As a minimum, the PPS Instrumentation and 480V Essential Bus Undervoltage Protection channels identified in Part 2 of Tables 4.4-1 through 4.4-5 shall be OPERABLE with TRIP SETPOINTS equal to or greater than (or less than when indicated) the values displayed in Part 1 of Tables 4.4-1 through 4.4-5.

l APPLICABILITY: As shown in Part 2 of Tables 4.4-1 through 4.4-5.

l ACTION:

With a PPS Instrumentation Functional Unit "as found" value less conservative than the value shown in the ALLOWABLE VALUE column of Tables 4.4-1 through 4.4-5, declare the channel inoperable, and apply the applicable ACTION statement requirement.

A channel that has been placed in the tripped condition as a Q result of an ACTION requirement shall be considered OPERABLE for the purposes of Specification 4.0.6 relative to entry into other OPERATIONAL MODES.

SURVEILLANCE REQUIREMENT (S) lSR5.4.1 O

r Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.4-2 Ol Specification LC0 4.4-1 l Table 4.4-1 (Part 1)

INSTRUMENTATION SETPOINTS FOR PLANT PROTECTIVE SYSTEM, SCRAM l

TRIP ALLOWABLE N0. FUNCTIONAL UNIT SETP0 INT VALUE la. Manual Scram (Control Room) . Not Applicable Not Applicable Ib. Manual Scram (Outside Control Room) Not Applicable Not Applicable l2a. STARTUP Channel High 18.3E+04 cps 19.3E+04 cps 2b. Wide Range Channel Rate of Change- High 14.5dpm 14.5dpm 3a. Linear Channel-High Varies as a Varies as a Channels 3,4,5 Function of Function of (Neutron Flux) (b) Indicated Indicated Thermal Thermal

" **r (*) " **r (*)

O Varies as a Varies as a 3b. Linear Channel-High Channels 6,7,8 Function of Function of (Neutron Flux) (b) Indicated Indicated Thermal Thennal Power (a) Power (a)

4. Primary Coolant Moisture (b) a) High Level Monitor <60.5 degree F <60.5 degree F dewpoint Tewpoint b) Loop Monitor <20.4 degree F <20.4 degree F Tewpoint Tewpoint

5. Reheat' Steam Temperature

-High (b) 11055 degree F 11061degreeF O

Fort St.- Vrain #1 O Techaice1 SPecificatioas Amendment No.

Page 4.4-2a l

Specification LC0 4.4-1 l Table 4.4-1 (Part 1) (Continued)

INSTRUMENTATION SETPOINTS FOR PLANT PROTECTIVE SYSTEM, SCRAM l

TRIP ALLOWABLE N0. FUNCTIONAL UNIT SETPOINT VALUE

6. Primary Coolant Pressure <64.6 psi below <67.0 psi below

-Programmed Low (b) Wormal, Formal, programmed programmed with Circulator with Circulator Inlet Inlet Temperature. Temperature Upper TRIP per Figure SETPOINT 3.3-1. Upper

> 635.4 psia.

limit to produced trip at > 633 psii.

O 7. erimery Cooient eressure

-Programmed High (b)

<44 psi enove Wormal,

<47 psi eeove normal, programmed programmed with Circulator with Circulator Inlet Inlet Temperature. Temperature Upper TRIP per Figure SETP0 INT of 3.3-1. Upper

< 744 psia. limit to Iower TRIP produce trip SETPOINT of at < 747

-< 536 psia, psis. Lower l

limit to trip l

at i 539 psia.

l

8. Hot Reheat Header Pressure >44 psig

>44 psig

-Low

>1529 psig >1529 psig l9. Main Steam Pressure-Low l10. Plant Electrical System-Loss >278V(c) >278V(c)

11. Two Loop Trouble Not Applicable Not Applicable

, Scram Logic O 12. High Reactor Building

<165 degree F Temperature (Pipe Cavity) 1161degreeF

r Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.4-2b l

Specification LC0 4.4.1 l Table 4.4-1 (Part 2)

INSTRUMENT OPERATING REQUIREMENTS FOR PLANT PROTECTIVE SYSTEM, SCRAM TOTAL CHANNELS MINIMUM FUNCTIONAL NO. OF TO CHANNELS APPLICABLE NO. UNIT CHANNELS TRIP OPERABLE MODES ACTION la. Manual Scram 2 1 2 P,L,S/U 1 (Control Room) S/D,R(d)

Ib. Manual Scram 3 2. 2 P,L,S/U,R 2 (Outside (d)

Control Room) 2a. STARTUP 2 1 2 R(d) 3 Channel-High 2b. Wide 3 2 2 S/U 2 Range Channel C Rate of Change-High 3a. Linear 3 2 2 P,L,S/U 2 Channel-High Channels 3,4,5 3b. Linear 3 2 2 P,L,S/U 2 Channel-High Channels 6,7,8

4. Primary Coolant Moisture a) High Level 2 1(e) 1(f) P,L 4 Monitor (k) b)LoopMonitor 3 2(e) 2/ Loo P,L 4 l

5. Reheat 3 2 2(h) P,L 2 l Steam Temperature

-High O

l

r Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.4-2c l

Specification LC0 4.4.1 l Table 4.4-1 (Part 2)

INSTRUMENT OPERATING REQUIREMENTS FOR PLANT PROTECTIVE SY51LM, SCRAM TOTAL CHANNELS MINIMUM FUNCTIONAL N0. OF TO CHANNELS APPLICABLE NO. UNIT CHANNELS TRIP OPERABLE MODES ACTION

6. Primary 3 2 2(1)(j) P 2 Coolant Pressure

-Programmed low

7. Primary 3 2 2(i)(j) P,L 2 Coolant Pressure

-Programmed High

8. Hot Reheat 3 2 2 P 2 Q Header Pressure

-Low

9. Main Steam 3 2 2 P 2

- Pressure i -Low

10. Plant 3 2 2(c) P,L,S/U 2 i Electrical System-Loss i 11. Two Loop- 3 2 2 P,L,S/U 2 Trouble L Scram Logic I

12. High Reactor 3 2 2 P,L,S/U 2 J

i Building l Temperature

! (PipeCavity) 1 0

1 l

Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.4-2d l Table 4.4-1 (Continued)

TABLE NOTATION l

'(a) Curves specifying the Linear Channel High Neutron Flux TRIP SETPOINT limit and ALLOWABLE VALUE as a function of indicated power level, which account for neutron detector decalibration, shall-be_ established for each fuel cycle. The neutron detector decalibration curves and instructions shall be approved by the NFSC prior to each fuel cycle. The detector decalibration curves approved by the NFSC shall be forwarded within 30 days of approval to the Regional Administrator, Region IV, The Commission. See Tables 3.3-1 and 4.4-4 for related limits.

l (b) See _also Specification LSSS 3.3, Table 3.3-1.

(c) One channel consists of three undervoltage relays each monitoring a single phase of a 480 VAC essential bus. A channel trip will occur when two of the three undervoltage relays comprising that channel operate after a preset time delay indicating loss of bus voltage. Initiation of a scram requires two of the three undervoltage relays on two of the three 480 VAC essential buses to operate.

(d)The applicability for the REFUELING MODE is anytime the reactor

-Q scram is reset and the control rod drive system is capable of rod withdrawal.

(e)The trip of one primary coolant high level moisture monitor accompanied by trips of two of the three loop moisture monitors for either loop will cause full scram.

(f) A primary coolant dew point moisture monitor shall not be l considered OPERABLE unless the following conditions are met:

1) Reactor Power Range Minimum Sample Flow l

STARTUP to 2%* > 1 scc **/sec.

Greater than 2% to 5%* I 5 sec/sec.

Greater than 5% to 20% T15 sec/sec.

Greater than 20% to 35% T30 sec/sec.

Greater than 35% to 100% TSO sec/sec.

2) Minimum flow of item 1) is alarmed in the control room and the alarm is set in accordance with the power ranges specified.

l

• Required for use as specified by LCO 4.4.5.

• • sec - Standard cubic centimeters Ol

Fort St. Vrain #1 O

V Technical Specifications Amendment No.

Page 4.4-2e

3) The ambient temperatures indicated by both temporary thermocouples mounted on the- flow sensors in penetrations B1 and B3 are less than 185 degree F.

4) Fixed alarms of ->l scc /sec and <75 scc /sec are OPERABLE.

l(g)PermissibleBypassConditions:

l I. Any circulator buffer seal malfunction.

l II. Either loop hot reheat header high activity.

l III. As stated in LCO 4.9.2.

(h) Two thermocouples from each loop, total of four, constitute one channel. For each channel, two thermocouples must be OPERABLE in at least one operating loop for that channel to be considered OPERABLE.

(i) One OPERABLE helium circulator inlet thermocouple in an operating loop is required for the channel to be considered OPERABLE.

(j) For programmed trip functions, a channel is inoperable when the channel which provides the programmed TRIP SETP0 INT is inoperable.

.(k)In the event that testing of existing PLANT PROTECTIVE SYSTEM (PPS) moisture monitors or installation and' testing of new moisture monitors take place, input trip functions to the PPS which cause scram, loop shutdown, circulator trip and steam water dump should be disabled.

During the time that the PPS moisture monitor trips are disabled, l

an adequate number of observers in direct communication with the i reactor operator shall be positioned in the control room in the location of pertinent instrumentation. The observer (s) shall continuously monitor the primary coolant moisture levels indicated by at least two OPERABLE moisture monitors, PPS or Analytical Monitors, and the primary coolant pressure indications, and shall alert the reactor operator to any indicated moisture or pressure change. During the time in which

l. the trip functions are disabled the requirements of LC0's 4.2.10 and 4.2.11 shall be met and primary coolant shall not exceed a moisture concentration of 100 ppmv.

1O l

l l

i

Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.4-2f Ul Table 4_.4-1 (CONTINUED) l fj_TIONSTATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the MINIMUM CHANNELS OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least STARTUP within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and be in SHUTDOWN in the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 2 - With the numbcr of OPERABLE channels less than the TOTAL NUMBER OF CHANNELS, restore the inoperable channel (s) to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the channel (s) is not restored to OPERABLE status within that 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> either; (a) Place the inoperable channel in the tripped condition unless trip of the channel would cause a scram or (b) If trip of the channel would cause a scram, place the reactor in a MODE where the limit does not apply within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 3 - With the number of OPERABLE channels one less than the MINIMUM CHANNELS OPERABLE requirement, suspend all p operations involving control rod withdrawal, terminate v incore maintenance and any remote operated mechanisms shall be retracted and the opening through the PCRV closed

~

as soon as practicable per LC0 4.5.2 and terminated fuel manipulation and the fuel handling mechanism shall be retracted into the fuel handling machine and the isolation valve closed as soon as practicable per LC0 4.7.1.

ACTION 4 - With the number of OPERABLE channels less than the TOTAL NUMBER OF CHANNELS, restore the inoperable channel (s) to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the channel (s) is not restored to OPERABLE status within that 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> either; (a) Place the inoperable channel in the tripped condition unless trip of the channel would cause a scram or, (b) In the special case of an inoperable moisture monitor, the moisture monitor input trip functions to the PPS which cause scram, loop shutdown, circulator trip, and steam / water dump may be disabled for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

During the time that the PPS moisture monitor trips are disabled, an adequate number of observers in direct communication with the reactor operator shall be positioned in the control room in the location of pertinent instrumentation. The observer (s) shall continuously monitor the primary coolant moisture I] levels indicated by at least two OPERABLE moisture monitors, PPS or Analytical Monitors, and the primary coolant pressure indications, and shall alert the t

Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.4-2g reactor operator to any indicated moisture or pressure change. During the time in which the trip functions are disabled the requirements of LC0's 4.2.10 and 4.2.11 shall be met and primary coolant shall not exceed a moisture concentration of 100 ppmv or c) If trip of the channel would cause a scram, place the reactor in a MODE where the limit does not apply within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

O O

Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.4-3

'~

l Specification LC0 4.4.1 l Table 4.4-2 (Part 1)

INSTRUMENTATION TRIP SETPOINTS FOR THE PLANT PROTECTIVE SYSTEM, LOOP SHUTDOWN t

TRIP ALLOWABLE NO. FUNCTIONAL UNIT SETPOINT VALUE la. Steam Pipe Rupture Under PCRV, Loop 1 (1) < 8.68 VDC < 8.86 VDC lb. Steam Pipe Rupture Under PCRV, Loop 2 (1) < 8.68 VDC < 8.86 VDC Ic. Steam Pipe Rupture, North Pipe Cavity Loop 1 (1) < 8.68 VDC < 8.86 VDC Id. Steam Pipe Rupture, South Pipe Cavity Loop 1 (1) < 8.68 VDC < 8.86 VDC

] le. Steam Pipe Rupture, North Pipe Cavity Loop 2 (1) < 8.68 VDC < 8.86 VDC if.

SteamPipeRupture,(South Pipe Cavity Loop 2 1) < 8.68 VDC < 8.86 VDC 2a. High Pressure, Pipe Cavity (1) < 1.3" H2O < 1.3" H20 2b. High Temperature, Pipe Cavity (1) i125degreeF i125degreeF 2c. High Pressure, Under PCRV (1) < 1.3" H70 < 1.3" H20 2d. High Temperature, Under PCRV (1) i125degreef i125degreeF l 3a. Loop 1 Shutdown Logic Not Applicat,le Not Applicable l3b. Loop 2 Shutdowr. Logic Not Applicable Not Applicable 4a. Circulator 1A and IB Shutdown - Loop Shutdown Logic Not Applicuble Not Applicable O

Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.4-3a 0l Specification LC0 4.4.1 l Table 4.4-2 (Part 1)

INSTRUMENTATION TRIP SETP0INTS FOR THE PLANT PRUltCIIVE SYbitM, LUUP 5HUTDOWN TRIP ALLOWABLE l N0. FUNCTIONAL UNIT SETP0 INT val.UE 4b. Circulator 1C and ID Shutdown - Loop Shutdown Logic Not Applicable Not Applicable Sa. Steam Generator Penetration Overpressure Loop 1 1796psig 1796psig Sb. Steam Generator Penetration Overpressure Loop 2 1796psig 1796psig 6a. High Reheat Header < 3.2 mrem /hr < 3.2 mrem /hr n

V Activity, Loop i Kbove Above Background Background 6b. P1gh Reheat Header < 3.2 mrem /hr < 3.2 mrem /hr A.ctivity, Loop 2 Above Above Background Background 7a. Low Superheat Header Temperature, Loop 1 (m) > 798 degree F 1798degreeF 7b. Low Superheat Header Temperature, Loop 2(m) > 798 degree F 1798degreeF 7c. High Differential < 44.8 degree F

< 44.8 degree F Ten n rature Between

'mp1andloop2(m)

8. Primary Coolant Moisture a) High Level ----------------See Table 4.4-1---------------

Monitor b)LoopMonitor ----------------See Table 4.4-1---------------

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Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.4-3b l Specification LC0 4.4.1 l Table 4.4-2 (Part 2)

INSTRUMENT OPERATING REQUIREMENTS FOR PLANT PROTECTIVE 5Y51LM, LOOP 5HUIDOWN TOTAL CHANNELS MINIMUM FUNCTIONAL NO. OF TO CHANNELS APPLICABLE NO. UNIT CHANNELS TRIP OPERABLE MODES ACTION la. Steam Pipe 3 2 2(n) P.L.S/U 5 Rupture Under PCRV, Loop 1 (1) lb. Steam Pipe 3 2 2(n) P,L,S/U 5 Rupture Under PCRV, Loop 2 (1)

Ic. Steam Pipe 3 2 2 P,L,S/U 5 Rupture, North Pipe Cavity Loop 1 (1) id. Steam Pipe 2 2 P,L,S/U 5 Q' Rupture, South 3

Pipe Cavity Loop 1 (1) le. Steam Pipe 3 2 2 P,L,S/U 5 Rupture, North Pipe Cavity Loop 2 (1)

If. Steam Pipe 3 2 2 P,L,S/U 5 Rupture, South Pipe Cavity Loop 2 (1) 2a. High Presst:re, 3 2 2 P L.S/U 5 Pipe Cavity (1) 2b. High 3 2 2 P,L,S/U 5 Temperature, Pipe Cavity (1) 2c. High Pressure, 3 2 2 P,L,S/U 5 Under PCRV (1)

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Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.4-3c p

V l Specification LC0 4.4.1 l Table 4.4-2 (Part 2)

INSTRUMENT OPERATING REQUIREMENTS FOR PLANT PROTECTIVE SYSTEM, LOOP 5HUTDOWN TOTAL CHANNELS MINIMUM FUNCTIONAL NO. OF TO CHANNELS APPLICABLE NO. UNIT CHANNELS TRIP OPERABLE MODES ACTION 2d. High 3 2 2 P,L,S/U 5 Temperature, Under PCRV (1) 3a. Loop 1 2 1 2 P,L,S/U, 6 Shutdown (o)

Logic 3b. Loop 2 2 1 2 P,L,S/U, 6 Shutdown (o)

Logic 4a. Circulator IA 2 1 2 P.L.S/U, 7 O and IB Shutdown

-Loop Shutdown (o)

Logic 4b. Circulator 1C 2 1 2 P,L,S/U, 7 and ID Shutdown (o)

-Loop Shutdown Logic Sa. Steam Generator 3 2 2 P,L,S/U 5 Penetration Overpressure Loop 1 Sb. Steam Generator 3 2 2 P,L,S/U 5 Penetration Overpressure Loop 2 6a. High Reheat 3 2 2 P,L,S/U 5 Header Activity, Loop 1 6b. High Reheat 3 2 2 P L.S/U 5 Header Activity, Loop 2

Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.4-3d q

U l Specification LC0 4.4.1 l Table 4.4-2 (Part 2)

INSTRUMENT OPERATING REQUIREMENTS FOR PLANT PROTECTIVE SYSTEM, LOOP 5HUTDOWN TOTAL CHANNELS MINIMUM FUNCTIONAL N0. OF TO CHANNELS APPLICABLE NO. UNIT CHANNELS TRIP OPERABLE MODES ACTION 7a. Low Superheat 3 2 2 P 5 Header Tempera-ture, Loop 1 (m) 7b. Low Superheat 3 2 2 P 5 Header Tempera-ture, Loop 2 (m) 7c. High 3 2 2 P 5 Differential Temperature Between Loop 1 and Loop 2 (m)

O 8. Primary Coolant Moisture a)HighLevel -----------------See Table 4.4-1----------------

Monitor l b)LoopMonitor -----------------See Table 4.4-1----------------

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Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.4-3e l Table 4.4-2 (CONTINUED) l TABLE NOTATION (1) Indication of Steam Pipe Rupture Under PCRV Loop 1 (Loop 2) or Steam Pipe Rupture North Pipe Cavity Loop 1 (Loop 2) or Steaiii Pipe Rupture South Pipe Cavity Loop 1 (Loop 2) and indica ~ tion of High Pressure-Pipe Cavity or High Temperature-Pipe Cavity or High Pressure-Under PCRV orHTghTemperature-UnderPCRVarerWuired for Loop 1 (Loop 2) Shiitdown.

(m) Low Superheat Header Temperature Loop 1 (Loop 2) must be accompanied by High Differential Temperature between Loop 1 and Loop 2 for Loop 1 (Loop 2) Shutdown.

(n) Each channel has 2 microphones connected in parallel with an ultrasonic amplifier. For the channel to be considered OPERABLE, both microphones and the amplifier must be OPERABLE.

(o) Applicable logic for each MODE depends upon the applicable Functional Unit parameters for the MODE and which circulator, circulator drive and loop are operating or are required to be OPERABLE.

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Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.4-3f l Table 4.4-2 (CONTINUED) l ACTION STATEMENTS ACTION 5 - With the number of OPERABLE channels less than the TOTAL NUMBER OF CHANNELS, restore the inoperable channel (s) to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the channel (s) is not restored to OPERABLE status within that 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, either; (a) Place the inoperable channel in the tripped condition unless trip of the channel would cause a loop shutdown 9.C (b) If trip of the channel would cause a loop shutdown, shut down the loop or place the reactor in a MODE where the limit does not apply within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 6 - With the number of OPERABLE channels one less than the MINIMUM CHANNELS OPERABLE requirement, restore the inoperable channel to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or shut down the affected loop within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTION 7 - With the number of OPERABLE channels one less than the O MINIMUM CHANNELS OPERABLE requirement, restore the inoperable channel to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or shut down the affected Helium Circulator (or loop) within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.4-4 l

Specification LC0 4.4.1 l Table 4.4-3 (PART 1)

INSTRUMENTATION TRIP SETPOINTS FOR THE PLANT PROTECTION SYSTEM, CIRCULATOR TRIP TRIP ALLOWABLE-NO. FUNCTIONAL UNIT SETPOINT VALUE

1. Manual Trip (Steam) Not Not Applicable Applicable

2. Circulator Speed - <11,495 rpm

~

<11,570 rpm

~

High(Steam)

3. Circulator Drain >8 psid

->8 psid Malfunction

4. Manual Trip (Water) Not Not Applicable Applicable

5. Circulator Speed - ~

<8,589 rpm -<8,670 rpm C High(Water)

6. Circulator Speed <1850 rpm Below <1974 rpm Below

- Low Programmed Nonnal As Normal As Programmed by Programmed by Feedwater Flow Feedwater Flow (4 circulators) (4 circulators) per Figure 4.4-la 7a. Loop 1, Fixed >177,500 lb/hr >171,750 lb/hr Feedwater T15.4% of normal T14.9% of normal

. Flow - Low (Both Full Load) Full Load) l Circulators) 7b. Loop 2, Fixed >177,500 lb/hr >171,750 lb/hr Feedwater T15.4% of normal T14.9% of normal Flow - Low (Both Full Load) Full Load) j Circulators) l l

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Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.4-4a p) t l Specification LC0 4.4.1 l Table 4.4-3 (PART 1)

INSTRUMENTATION TRIP SETPOINTS FOR THE PLANT PROTECTION SYSTEM, CIRCULATOR TRIP TRIP ALLOWABLE NO. FUNCTIONAL UNIT SETPOINT VALUE 8a. Loop 1, Programed <211,000 lb/hr <230,530 lb/hr Feedwater Flow - T18.3%)Below T20%) Below Low (Both normal as normal as Circulators) programmed by programed by Circulator Circulator Speed. (4 Speed.(4 circulators) circulators) per Figure 4.4-la 8b. Loop 1, Programed <211,000 lb/hr <230,530 lb/hr Feedwater Flow - T18.3%)Below T20%)Below Low (OneCirculator) normal as normal as programmed by programed by (v~') Circulator Circulator Speed. (2 Speed. (2 circulators) circulators) per Figure 4.4-lb 9a. Loop 2, Programmed <211,000 lb/hr <230,530 lb/hr Feedwater Flow - T18.3%)Below T20%) Below Low (Both normal as normal as Circulators) programmed by programmed by Circulator Circulator Speed.(4 Speed.(4 circulators) circulators) per Figure 4.4-la 9b. Loop 2, Programmed <211,000 lb/hr <230,530 lb/hr Feedwater Flow - T18.3%) Below T20%)Below Low (OneCirculator) nonnal as normal as programmed by programmed by Circulator Circulator Speed.(2 Speed (2 circulators) circulators) per Figure o 4.4-Ib L]

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Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.4-4b Ol Specification LCO 4.4.1

'l Table 4.4-3 (PART 1)

INSTRUMENTATION TRIP SETPOINTS FOR THE PLANT PROTECTION SYSTEM, CIRCULATOR TRIP TRIP ALLOWABLE NO. FUNCTIONAL UNIT SETPOINT VALUE 10a. Circulator Seal ->-5.2" H20, >-6" H20, Malfunction-Low 10b. Circulator Seal ~

<+74.8" H2O <+75.6" H20

~

Malfunction-High

11. Loss of Circulator ->459 psid >459 psid Bearing Water

12. Circulator -

<796 psig -

<796 psig Penetration Overpressure O

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Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.4-4c l Specification LCO 4.4.1 l Table 4.4-3 (Part2)

INSTRUMENT OPERATING REQUIREMENTS FOR PLANT PROTECTIVE SY5TtM, CIRCULATOR TRIP TOTAL CHANNELS MINIMUM FUNCTIONAL NO. OF TO CHANNELS APPLICABLE NO. UNIT CHANNELS TRIP OPERABLE MODES ACTION

1. Manual Trip 1 1 1 All 8 (Steam) Modes (r)

2. Circulator 3 2 2 All 8 Speed-High Modes (r)

(Steam)

3. Circulator 3 2 2 P,L(S/U, 8

Drain S/Ds)

Malfunction

4. Manual Trip 1 1 1 All 8 (Water) Modes (r)

O S. Circulator 3 2 2 All 9 Speed-High Modes (r)

(Water)

6. Circulator 3 2 2 P 8 Speed-Low Progranned 7a. Loop 1, 3 2 2 P 8 Fixed Feedwater Flow-Low 7b. Loop 2, 3 2 2 P 8 Fixed Feedwater Flow-Low 8a. Loop 1, 3 2 2 P 8 Programmed Feedwater Flow-Low (Both Circulators)

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Fort St. Vrain #1 Technical Specifications >

Amendment No.

Page 4.4-4d l Specification LCO 4.4.1 l Table 4.4-3 (Part 2)

INSTRUMENT OPERATING REQUIREMENTS FOR PLANT PROTECTIVE 5Y51tM, CIRCULATUR TRIP TOTAL CHANNE LS MINIMUM FUNCTIONAL NO. OF TO CHANNELS APPLICABLE NO. UNIT CHANNELS TRIP OPERABLE MODES ACTION 8b. Loop 1, 3 2 2 P 8 Programmed Feedwater Flow-Low (One Circulator)

Sa. Loop 2, 3 2 2 P 8 Programmed Feedwater Flow-Low (Both Circulators) 9b. Loop 2, 3 2 2 P 8 O ero9r ed Feedwater Flow-Low (One Circulator) 10a. Circulator 3 2 2(t) All 8 Seal Modes (r)

Malfunction-Low 10b. Circulator 3 2 2(t) All 8 Seal Modes (r)

Malfunction-High

11. Loss of 3 2 2 All 8 Circulator Modes (r)

Bearing Water

12. Circulator 3 2 2 All 8 Penetration Modes (r)

Overpressure O

Fort St. Vrain g Technical Specifications l Q Amendment No.

Page 4.4-4e

-125 % -

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SPEED LOW li !it! !O OPERATION Uli Uillifil M ALLOWABLE J Nil !!!-ill iDI il4 UPPER LIMIT

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r1 li ,IN lih. Idi it!! iil.i i. IfTUI ttfi ttti t TTITTit!! ifil fli !! 9660 RPM I~[fi!!! SPEED HIGH i

~

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Ellif ilii :ii liii Yi Hijilii !!!i11PIisih!i' ' fiji }!!! if'it PROGRAMMED?! fi{i !!!!

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" 4 LOWER LIMIT i !70! ? hMifE  !!il ild i m.

l*/o.t\em ALLOWABLE min .m ..." . . - . . . . . . . . . . . .

.N. . .ii. !!. ""

.@.rif li l!ti ti! FEEDWATER ! !Iii lii' E!fEli iHi f2

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.n.r: n ". 288,170 LB/HRj ;at - n .". .a

- 250 ," n"

"..N. ..!.! !!!- Ii! !!!!lii! IIll i' '.!!!! illi

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iH H!i *t-li if! 4200 RPM t ".

.n PROGRAMMED a - . a. .- ..a .- ~. -

'.h ' " . .II!U i IIUIN. "'. . e'

-15% - . '

l h "Ji liH #Ii! illi LIE FEEDWATER ".

FIXED FEEDWATER"":"'3:SS "" T

!' " . ^ i"" "ti"" '" "j:: ""# """ li"" iiL{E EFLOW LOW ALLOWABLE tjilli!! FLOW LOW

";g;;;;i j;;

VALUE (171,750 L8/HR) j! ! ;W jog (IMITgj jq ;;;; g; jj;; ;;;; ;;g ;;j pi,3;;g ;;p, , ,

-0 , ,

-0% i ,

4000 0000 8000 10,000 0 2000 CIRCULATOR SPEED, RPM Figure 4.4-la FEEDWATER FLOW vs. CIRCULATOR SPEED FOR OPERATION IN THE POWER MODE g (4 CIRCULATORS)

Fort St. Vrain Technical Specifications Amendment No.

G 4.4-4f b

'l

.I

wm .

m

-125 % i jiih. n.i -m..: .ml"li..2:rt n gn.-.);.siintiD$j;n;h-p_

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- 1250 .... .; .. .

n  :.i.!F...Pa

. . . a- t..E, .i. l.. t.t..H.1.'570.-.R P.M. ). .."n'~i a.t r.t. d.i _ F:Hir t."...

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< $ 4f +i ^ " FEEDWATER -n E -600 # - ~ "" "~ " ..

a r tiii li: !E" SPEED LOW "E" k-MM"' +H"{!~ *! "!"!  ! !i' - . ' ' - FLOW LOW H Wt h

• LOWER LIMIT E U- 1.:

_t Ata +ttt <tt".tt~.. .t:/ .-iff.h... I.ll.i"lU. .I!.U..I.!!!II!L..I. i U..li y .t$.!.p !!.E.. !.@ ..i.(7. P_P.E.R L.iMIT.

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• " * :it:'

• " " de

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FEEDWATER

.1j. i.ji .

. . . .. I _jjij. jjg gj j[ . 8600 jjj!ity . . . .. __.._r_._ RPy . ,;,,, ,,3pf!y[((j";g FLOW,,L,,O,W , di !!I! Nil !!!! i li !!!.' .#ili , , , , , , , , .

                                                                                                                                                                                                                        ,!!,!!.If!,i ,i:i "I, Ii!i! l,i li!!,, I,it,, "Il !'I!., !!!i :ih_

ni' yti i : it PROGRAMMED"

      -15%                                 -   -             * * "                     -

i FIXED FEEDWATER I FEEDWATER jg igi jjji igj jgy [j gy j!ii pp gy gj gp FLOW LOW ALLOWA8LE i: !ti! 5~2 FLOW LOW  ! VAL.UE.4,.171..,.750.LB./ ( HR) .h

t. . .r,tt:. .i.:i.i..:t) LOWER LIMIT g ~~ ..

                                                                                                                                                                                                                                     ,t : L;g gg;g. . t, -
                                                                                                                                                                                                                                                                                           .m    t..

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                                                                                                                                                                                                                                                                                                           ..~    ..
      -0%            -0                                                                         - - , .                                                                                      i                                              ,

0 2000 4000 0000 8000 10,000 CIRCULATOR SPEED, RPM , Figure 4.4-Ib FEEDWATER FLOW vs. CIRCULATOR SPEED FOR OPERATION IN THE POWER MODE (2 CIRCULATORS) m

Fort St. Vrain #1 Technical Specifications Amendment No. Page 4.4-49 l Table 4.4-3 (Continued) l TABLE NOTATION (r) Applicable in MODES where the circulator is required to be OPERABLE per LC0 4.2.1. and LC0 4.2.2. (s) Applicable when the helium circulator (s) is being driven by steam. (t) Circulator seal malfunction channels can be bypassed when the opposite loop is shut down -or when the circulator seal malfunction is tripped on the atTer circulator in the same loop. O O

                            .                  _ _ _ _              ..y.                            . _      - _ _ _ _ . _ _ - . _ _ - _ __ - .

7 t ,, Fort St. Vrain #1 s Technical Specifications Amendment No. Page 4.4 ah (} ,

                                                                                ;              3.

l3 Table 4.4 4 (CONTINUED) J. N lg ACTION STATEMENTS ACTION 8 - With the number of OPE 9ABLE channels less than the TOTAL s NUMBER OF CHANNELS, rcstore the inoperable channel (s) to r- U OPERABLE status within,12 hours. If the channel (s) is not s restored to OPERABLF qtatus within that 12 hours, either;

                     .it                   ,
                                             ,                  ,        :p                                                                     ;

(a) Place the fnoperable channel in the tripped condition

                   ,          i                                     of the channel would cause a helium l

unless circulator trip (or loop) shistdown or (o) If trip of ths criannel would Unuse a helium circulator

   ,                                   (or, loop), shutdown , shut dcwn the 'affected helium circulator (or loop) or piace the rea: tor in a MODE where'the limit does not apply within the next 12 hours.

ACTION 9 - With the number of OPERABLE channels less than the TOTAL NUMBER OF CHANNELS, restcre the inoperable channel (s) to i OPERABLE status within 12 hmrs. If the charinel(s) is not restored to OPERABLE status hithin that 12^ hours, either; O (a) Place the inoperable cP4nnel in the tripped condition unless trip of the channel would cause a helium circulator (orloop)shotdownor (b) If trip of the channel would cause a helium circulator (or loop) shutdown to occur, shut. down the affected helium circulator. (or loop), or declare the water turbine drive inoperable'and comply with LC0 4.2,1 and 4.2.2 and J (c) If the TOTAL NUMBER 0F CtlANNELS4 1s not maintained OPERABLE on at laast one circulator per loop, the reactor shall be' reduced to 50% of RATED THERMAL POWER within 12 hours of discovery 4 e 4 t, k h 3 #,

Fort St. Vrain #1 Technical Specifications Acendment No. eg Page 4.4-5 Ul Specification LCO 4.4.1 l Table 4.4-4 (Part1) INSTRUMENTATION TRIP SETPOINTS FOR THE PLANT PROTECTIVE SYSTEM, ROD WITHDRAWAL PROHIBIT (RWP) TRIP ALLOWABLE NO. FUNCTIONAL UNIT SETPOINT VALUE

1. STARTUP channel-Low -> 4.2 cps > 3.2 cps Count rate (Channels 1 and 2) 2a. Linear Channel-5% RWP* < 5% (x)

                                                  -                -< 5%

(Channels 3, 4 and 5) 2b. Linear Channel-5% RWP* < 5% (x)

                                                  --               -< 5%

(Channels 6, 7 and 8) 3a. Linear Channel-30% RWP* -<30%(y) -< 30% (Channels 3, 4 and 5) O 3b. Linear Channel-30% RWP* (Channels 6, 7 and 8)

                                                  -<30%(y)          -< 30%

4a. Wide Range Channel Rate of -

                                                   < 1.5 dpm        -< 2 dpm Change - High (Channels 3, 4 and 5) 4b. STARTUP Channels Rate of             i1.5dpm           i 2 dpm Change - High (Channels 1 and 2)

Sa. Linear Channel-High power RWP Varies as a Varies as a (Channels 3, 4 and 5) Function of Function of I Indicated Indicated Thermal Thermal l Power (z) Power (z) l i I l * % of RATED THERMAL POWER l l () l l [

Fort St. Vrain #1 Technical Specifications Amendment No. Page 4.4-Sa l Specification LCO 4.4.1 l Table 4.4-4 (Part1) INSTRUMENTATION TRIP SETPOINTS FOR THE PLANT PROTECTIVE SYSTEM, FOR ROD WITHDRAWAL PROHIBIT (RWP) TRIP ALLOWABLE NO. FUNCTIONAL UNIT SETPOINT VALUE Sb. Linear Channel-High power RWP Varies as a Varies as a (Channels 6, 7 and 8) Function of Function of Indicated Indicated . Thermal Thermal Power (z) Power (z)

6. Multiple Rod Not Not Pair Withdrawal Applicable Applicable O

Vg

Fort St. Vrain #1 Technical Specifications Amendment No. Page 4.4-5b [] l Specification LC0 4.4.1 l Table 4.4-4 (Part 2) INSTRUMENT OPERATING REQUIREMENTS FOR THE PLANT PROTECTIVE SYSTEM, ROD WITHDRAWAL PROHIBIT (RWP) TOTAL CHANNELS MINIMUM FUNCTIONAL NO. OF TO CHANNELS APPLICABLE NO. UNIT CHANNELS TRIP OPERABLE MODES ACTION

1. STARTUP 2 1 2 S/U,R(aa) 10 Channel-Low (bb)

Count Rate (Channels 1-and 2) 2a. Linear 3 2 2 S/U(x) 11 Channel-5% RWP (Channels 3,4 and 5) 2b. Linear 3 2 2 S/U(x) 11 O Channei-5% RWP (Channels 6, 7 and8) 3a. Linear 3 2 2 L(y) 11 Channel-30% RWP (Channels 3, 4 and 5)

;      3b. Linear               3           2         2       L(y)        11 Channel-30% RWP (Channels 6, 7 and8) 4a. Wide Range            3          2         2       S/U         11 l             Channel Rate l            of Change-High (Channels 3,4 and 5) 4b. STARTUP               2           1        2      'S/U(bb)      10 Channels Rate f)         of Change-High v          (Channels 1 and 2) i

Fort St. Vrain #1 Technical Specifications Amendment No. Page 4.4-5c l Specification LC0 4.4.1 l Table 4.4-4 (Part 2) INSTRUMENT OPERATING REQUIREMENTS FOR THE PLANT PROTECTIVE SYSTEM, ROD WITHDRAWAL PROHIBIT (RWP) TOTAL CHANNELS MINIMUM FUNCTIONAL N0. OF TO CHANNELS APPLICABLE NO. UNIT CHANNELS TRIP OPERABLE MODES ACTION Sa. Linear 3 2 2 P,L,S/U 11 Channel-High Power RWP (Channels 3, 4 and 5) Sb. Linear 3 2 2 P,L,S/U 11 Channel-High Power RWP (Channels 6, 7 and 8)

6. Multiple Rod ------------------Not Applicable-----------------

O Pair Withdrawal f 4 O i

Fort St. Vrain #1 Technical Specifications Amendment No.

 ,s Page 4.4-5d

(# I l Table 4.4-4 (Continued) TABLE NOTATION l (x) The icw power RWP bistable automatically resets at approximately 4% of RATED THERMAL POWER after reactor power is decreased from greater than 5% of RATED THERMAL POWER. (y) The Power range RWP bistables automatically reset at approximately 10% of RATED THERMAL POWER after reactor power is decreased from greater than 30% of RATED THERMAL POWER. The RWP may be manually reset between approximately 10% and 30% of RATED THERMAL POWER. (z) Curves specifying the Linear Channel High Neutron Flux TRIP SETPOINT limit and ALLOWABLE VALUE as a function of indicated power level, which. account for neutron detector decalibration, shall be established for each fuel cycle. The neutron detector decalibration curves shall be approved by the NFSC prior to each fuel cycle. The detector decalibration curves approved by NFSC shall be forwarded within 30 days of approval to the Regional Administrator; Region IV, The Commission. See Tables 4.4-1 and 3.3-1 for related limits. D (aa') The applicability for the REFUELING MODE is anytime the reactor h scram is reset and the control rod drive system is capable of rod withdrawal. (bb) The STARTUP channel may be disabled above E-03% of RATED THERMAL POWER. l I (a

Fort St. Vrain #1 Technical Specifications Amendment No. Page 4.4-Se l . Table 4.4-4 (CONTINUED) l ACTION STATEMENTS ACTION 10 - With the number of OPERABLE channels one less than the MINIMUM CHANNELS OPERABLE requirement; (a) Suspend all operations involving control rod withdrawal, terminate incore maintenance and any remote operated mechanisms shall be retracted and the opening through the PCRV closed as soon as practicable per LC0 4.5.2 and terminate fuel manipulation and the the fuel handling mechanism shall be retracted into the fuel handling machine and the isolation valve closed as soon as practicable per LC0 4.7.1 and (b) Either restore the inoperable channel to OPERABLE status within 12 hours, or actuate the Rod Withdrawal Prohibit, or be in at East STARTUP within the next 12 hours anTbe in SHUTDOWN in the following 12 hours. (3 ACTION 11 - With the number of OPERABLE channels less than the TOTAL v NUMBER OF CHANNELS, restore the inoperable channel (s) to OPERABLE status within 12 hours or suspend all operations involving control rod withdrawa'1, terminate incore maintenance and any remote operated mechanisms shall be retracted and the opening through the PCRV closed as soon as practicable per LC0 4.5.2 and terminate fuel manipulation and the fuel handling mechanism shall be retracted into the fuel handling machine and the isolation valve closed as soon as practicable per LC0 4.7.1.

Fort St. Vrain #1 Technical Specifications Amendment No. Page 4.4-6 l Specification LC0 4.4.1 l Table 4.4-5 (Part 1) TRIP SETP0INTS FOR 480V AC ESSENTIAL BUS UNDERVOLTAGE PROTECTION TRIP TIME ALLOWABLE l N0. FUNCTIONAL UNIT SETPOINT DELAY VALUE

1. Plant Electrical > 278V 28.5 - 31.5 > 278V System - Loss Tee also seconds -( 35 (Scram) Table 4.4-1 - seconds

2. Degraded Voltage 396V - 436V 115 - 125 Not seconds Applicable

3. Loss of Voltage- 361V - 383V CV-2 Relay Not Automatic Throw Setting of Applicable Over(ATO) Time Dial 5

4. Loss of Voltage-- 318V - 338V CV-2 Relay Not D.G. Start, Load Setting of Applicable Time Dial 6 Q Shed and Load Sequence i

O

Fort St. Vrain #1 Technical Specifications Amendment No. Page 4.4-6a l Specification LC0 4.4.1 l Table 4.4-5 (Part 2) INSTRUMENT OPERATING REQUIREMENTS FOR 480 VAC ESSENTIAL BUS UNDERVOLTAGE PROTECTION TOTAL CHANNELS MINIMUM FUNCTIONAL N0. OF -TO CHANNELS APPLICABLE NO. UNIT CHANNELS TRIP OPERABLE MODES ACTION

1. Plant Electrical --------------See Table 4.4-1--------------

System - Loss (Scram) j ' 2. Degraded Voltage 3 2 2(dd) All Modes '2 1

    ' 3. Loss of Voltage-       3          2       2(dd)     All Modes    12 Automatic Throw-over (ATO)

4. Loss of Voltage- 3 2 2(ee) All Modes 12 D.G. Start, Load Shed and Load Sequence

 .Q 1

i l O l l l 6

Fort St. Vrain #1 Technical Specifications Amendment No. Page 4.4-6b l Table 4.4-5 (Continued) l TABLE NOTATION (dd)A channel consists of one relay monitoring a single phase of a 480 VAC essential bus. The protective action occurs when two of three relays are activated. (ee)One channel consists of three undervoltage relays each monitoring a single phase of a 480 VAC essential bus. A channel trip will occur when two of the three undervoltage relays comprising that channel operate after a preset time delay indicating loss of bus voltage. Initiation of a scram requires two of the three undervoltage relays on two of the three 480 VAC essential buses to operate. O O

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Fort St. Vrain #1 Technical Specifications Amendment No. Page 4.4-6c 'O l Table 4.4-5 (CONTINUED) l ACTION STATEMENTS ACTION 12 - With the number of OPERABLE channels less than the TOTAL NUMBER OF CHANNELS, restore the inoperable channel (s) to OPERABLE status within 12 hours. If the channel (s) is not restored to OPERABLE status within that 12 hours, either; (a) Place the inoperable channel in the tripped condition unless trip of the channel would cause the protective action to occur g (b) If trip of the channel would cause the protective action to occur, declare the affected 480 VAC essential bus inoperable and comply with the requirements of LC0 4.6.1.1 or LC0 4.6.1.2. O

m Fort St. Vrain #1 Technical Specifications Amendment No. Page 4.4-7 NOTE l The Bases contained in succeeding pages summarizes the reasons for the Specifications in Section 4.4.1, but in accordance with 10CFR50.36 are not part of these Technical Specifications. O LO

Fort St. Vrain #1 Technical Specifications Amendment No. Page 4.4-8 lBasesforSpecificationLC04.4.1 The PPS automatically initiates protective functions to prevent established limits from being exceeded. In addition, other protective instrumentation is provided to initiate action which mitigates the consequences of accidents. This specification provides the limiting conditions- for operation necessary to preserve the effectiveness of these instrument systems. l The TRIP SETPOIN1S are included in this section of the Specification. The bases' for these settings are briefly discussed below. Additional discussions pertaining to the scram, loop shutdown and circulator trip inputs may be found in Sections 7.1.2.3, 7.1.2.4, and 7.1.2.6, respectively, of the FSAR. High moisture instrumentation is discussed in FSAR Section 7.3.2. a) Scram Inputs The . simultaneous insertion of the control rods will be initiated by the following conditions: lManualScram A Manual Scram is provided to give the operator means for .,O emergency shutdown of the reactor independent of the automatic reactor protective system. The Reactor Mode Switch (RMS) in the "off" position also causes a manual scram. lSTARTUPChannel-HighCountRate High start-up count rate is provided as a scram for use during fuel loading and preoperational testing. lLinearChannel-High(NeutronFlux) l See Technical Specification LSSS 3.3 lWideRangeChannel-RateofChange-High High rate of change of neutron flux is used as a scram input during plant start-up and results in additional protection to the Linear Channel - High scram in case of accidental control rod withdrawal . The TRIP SETP0 INT is selected to be above the operating rate of flux change. This scram TRIP SETPOINT is active only when the Interlock Sequence Switch is in STARTUP position. l Primary Coolant Moisture - High l See Technical Specification LSSS 3.3. O

Fort St. Vrain #1 Technical Specifications Amendment No. p Page 4.4-9 d lReheatSteamTemperature-High I l See Technical Specification LSSS 3.3 l Primary Coolant Pressure -- Programmed Low l See Technical Specification LSSS 3.3. lPrimaryCoolantPressure-ProgrammedHigh l See Technical Specification LSSS'3.3. lHotReheatHeaderPressure-Low Low reheat steam pressure is an indication of either a cold reheat steam line or a hot reheat steam line rupture in a section of line common to both loops. Loss of the cold reheat steam line results in loss of the steam supply to the circulators which necessitates plant shutdown. The direct scram in this case precedes a scram resulting from the two-loop trouble. The loss of either steam line results in loss of plant generation output, and a reactor scram is appropriate in this situation. The- TRIP SETP0 INT is selected to be below normal operating and transient levels, which vary over a wide range. O l Main Steam Pressure - Low Low main steam pressure is an indication of main steam line rupture or loss of feedwater flow. Immediate shutdown of the reactor is appropriate in such a situation. In addition, the superheater outlet stop check values are automatically closed to reroute main steam to the flash tank (through the individual loop bypass valves anddesuperheaters). This is necessary for the continued operation of the helium circulators on steam. The TRIP SETP0 INT is selected to be below normal operating levels and system transients. Plant Electrical System Power-Loss Loss of plant electrical system power requires a scram to prevent any power-to-flow mismatches from occurring. A preset time delay is following a power loss before the scram is initiated to lprovided allow an emergency diesel generator to start. If it does start, the scram is avoided. lTwo-LoopTroubleScramLogic Operation on one loop at a maximum. of about 50% power may continue following the shut down of the other loop (unless preceded by scram as in the case of high moisture). Onset of trouble in the o remaining loop (two-loop trouble) results in a scram. Trouble is () defined as a signal which normally initiates a loop shutdown. Simultaneous shutdown signals to both loops results in shut down of one of the two loops only, and a reactor scram.

Fort St. Vrain #1 Technical Specifications Amendment No. O Page 4.4-10 O lHighReactorBuildingTemperature,PipeCavity

   ]       High temperature in the pipe cavity would indicate the presence of an undetected steam leak or the failure of the steam pipe rupture detection system to determine the loop in which the leak had occurred and to shut the faulty loop down.

l The TRIP SETPOINT has been set above the temperature that would be expected to occur in the pipe cavity if the steam leak were detected and the faulty loop shut down for all steam leaks except those of major proportion such as that due to an offset rupture of one of the steam lines. An undetected steam leak or pipe rupture under the PCRV within the support ring would also be detectable in the pipe cavi ty, therefore only one set of sensors and logic is required to monitor both areas. b) Loop Shutdown Inputs The following loop shutdown inputs are provided primarily for equipment protection and are not relied upon to protect Safety Limits. Malfunction of these items could prevent a scram due to g loss of the two loop trouble scram input. U l Steam Pipe Rupture In The Reactor Building The purpose of the ultrasonic detectors is to identify the specific secondary coolant loop within the reactor building containing a pipe rupture. Ultrasonic noise caused by escaping steam in conjunction with a pressure or temperature rise will cause the appropriate loop to shut'down. The trip of the ultrasonic detection system is set at a level lwhichcorrespondsto8.68VDCoutputfromthe ultrasonic amplifier. The pressure and temperature trips are set above normal operating building pressure and temperature levels. lHighPressure-PipeCavity The TRIP SETPOINT is above normal reactor building pressure of 0.25" w.g. but below the pressure of about- 3" w.g. at which the reactor building louvers open to relieve any overpressure condition. lHighTemperature-PipeCavity The TRIP SETPOINT is established to be above the normal ambient temperature in the pipe cavity, and low enough to assure a fast response to steam pipe ruptures in the pipe cavity. 1 b

Fort St. Vrain #1 Technical Specifications Amendment No. Page 4.4-11 lHighPressure, Under PCRV The TRIP SETPOINT is above normal reactor building pressure of 0.25" w.g. but below the pressure of 3" w.g. at which the reactor building louvers open to relieve any overpressure condition. lHighTemperature,UnderPCRV The TRIP SETPOINT is established to be above the normal temperature inside the PCRV support ring to reduce the number of spurious trips. The ambient temperature under the PCRV is normally higher than that in the pipe cavity. Conversely, the TRIP SETPOINT is low enough to provide a fast response in the event of a steam pipe rupture. lShutDownofBothCirculators(LoopShutdownLogic) l Shut down of both circulators is a loop shutdown input which is necessary to insure proper action of the reactor protective (scram) system through the two-loop trouble scram in the event of the loss of all circulators and low feedwater flow. lSteamGeneratorPenetration-Overpressure (Loop 1/ Loop 2) steam 9eaerator Penetration overpressure is iadicative of a P iPe O rupture within the penetration. A loop shutdown is appropriate for such an accident, and the helium pressurizing line to the penetration is closed to prevent moisture backflow to the purified helium system. The penetration overpressure is handled by relief valves; however, to minimize the amount of steam / water so released, the contents of the steam generator in that loop are also dumped. The steam generator interspace rupture discs are set at 825 psig (nominal). The burst pressure range (+2%) is 809 psig to 842 (Technical Specification LSSS 3.3, Table T.3-1). The relief valve is sized to allow a 370 psi pressure drop in a safety valve inlet line when the valve is relieving at nameplate capacity of 126,000 lb/hr superheated steam at 1000 degree F. This prevents the penetration pressure from exceeding the reference pressure of 845 psig. lReheatHeaderActivity-High(Loop 1/ Loop 2) High reheat- header activity is an indication of a reheater tube rupture resulting in leakage of reactor helium into the steam system. The TRIP SETPOINT assures detection of major reheat tube rupture and an on scale reading, with up to design value -circulating activity for post accident monitoring. Detection of smaller size leaks or leaks with low circulating coolant activity can be detected and alarmed by the backup reheat condensate monitors and/or the air ejector monitor. O l l l L.

Fort St. Vrain #1 Technical Specifications Amendment No. Page 4.4-12 lLowSuperheatHeaderTemperature(Loop 1/ Loop 2andDifferential) Low superheat header temperature in a loop is indicative either of a feedwater valve or controller failure yielding an excessive loop feedwater flow or a deficiency of helium flow, and a loop shutdown is in order. The required coincident high differential temperature between loops functions to prevent the loop trip from occuring during normal operation at low main steam temperatures such as in a normal plant shutdown. c) Circulator Shutdown Inputs All circulator shutdown inputs (except circulator speed high on water turbines) are equipment protection items which are tied to two loop trouble through the loop shutdown system. These items are included in Table 4.4-3 because a malfunction could prevent a scram due to . loss of the .two loop trouble scram input. l Circulator speed high on. water turbines is included to assure continued core cooling capability on loss of steam drive. lCirculatorSpeed-Low 4 Too low a circulator speed causes a mismatch between thermal 4 power input and . heat removal (feedwater flow rate) in . a steam " O generator, which may result in flooding the superheater section. The circulator trip causes an automatic adjustment, as required, in the

turbine governor setting, feedwater flow rate, and remaining circulator speed to maintain stable steam pressure and temperature conditions.

l Loop 1/ Loop 2FixedFeedwaterFlow-Low The Fixed Feedwater Flow - Low is an equipment protection feature l designed to protect the steam generator from overheating due to complete loss of feedwater flow. l l Loop 1/ Loop 2ProgrammedFeedwaterFlow-Low I' A Programmed Feedwater Flow - Low prevents prolonged operation in the region of speed versus_ flow which may cause excessive superheat steam temperatures. r lLossofCirculitorBearingWater In order to prevent circulator damage upon loss of normal and backup bearing water supplies, a gas pressurized water accumulator is fired when water pressure falls below the TRIP SETP0 INT value. The TRIP SETPOINT value is selected so that adequate water pressure is available during -circulator coastdown, which lasts for about 30 seconds, to maintain clearances within the circulator bearings of at O least 0.001 in. Tests and analyses have shown that a trip at 450 psid provides substantial clearance margin above 0.001 in. when the circulators are operating at normal speeds.

    ..n,,

Fort.St. Vrain #1 Technical Specifications Amendment No. Page 4.4-13 l Circulator Penetration Trouble , Circulator penetration overpressure is indicative of a pipe rupture within the penetration. A circulator trip is appropriate for such an accident and the helium pressurizing line to the penetration

      -is closed to prevent moisture backflow to the purified helium system.
      !The overpressure is handled by the penetration relief valves. The penetration interspace rupture discs are set at 825 psig (nominal).

The burst pressure range (+ 2%) is 809 psig to 842 psig (Technical Specification LSSS 3.3 Table T.3-1). The relief valve is sized to allow a 40 psi pressure drop in the safet valve is relieving at nameplate capacity (y valve inlet line when the 170gpm). l Circulator Drain - Malfunction This trip.is provided to prevent steam from entering the bearing of an operating circulator. A differential pressure controller is utilized to maintain the bearing water main drain pressure above the steam turbine exhaust pressure. When the pressure differential drops, the steam water drain control valves are opened to prevent steam from entering the bearings. If the above controls do not work, three PPS differential pressure switches for each circulator, set at

       > 8 psid, will initiate an automatic shut down of the circulator.

C l Circulator Speed - High (Steam) The speed sensing system response and trip setting are chosen so that under the maximum overspeed situation possible (loss of restraining torque) the circulator will remain within design capabilities. lCirculatorTrip-Manual (Steam / Water) A manual trip of each circulator for both steam and water turbine drives is available so that in an emergency an operator can trip a circulator when necessary. l Circulator Seal - Malfuncton (Low /High) A high reverse differential of -6" H2O would be reasonable evidence that bearing water is leaking into the primary coolant system. An increasing differential pressure of +75.5" H2O would be reasonable evidence that primary coolant is leaking into the bearing water and thus into the closed circulator service system. In both cases a circulator trip with brake and seals set is appropriate. lCirculatorSpeed-High(Water) The TRIP SETPOINT has been established above normal operating speed. Equipment testing assures that this TRIP SETPOINT will prevent failure due to fatigue cracking.

Fort St. Vrain #1 Technical Specifications Amendment No. Page 4.4-13a O-l d) Rod Withdrawal Prohibit Inputs The termination of control rod withdrawal to prevent further reactivity addition will occur with the following conditions: lSTARTUPChannel-LowCountRate l STARTUP Channel-Low Count Rate is provided to prevent control rod withdrawal and . reactor startup without adequate neutron flux indication. The trip level is selected to be above the background noise level. lLinearChannel-5%RWP l Linear Channel (5% Power) directs the operator's attention to

   . either a downscale failure of a power range channel or improper lpositioningoftheInterlockSequenceSwitch.

l Linear Channel - 30% RWP Linear Channel (30% Power) is provided to prevent control rod withdrawal if reactor power exceeds the Interlock Sequence Switch limit for the " Low Power" position. Q j STARTUP Channel / Wide Range Channel-Rate of Change - High High Rate of change of neutron flux on the wide range channels (12 dpm) initiates an RWP. lLinearChannel-HighPowerRWP High neutron flux level from the power range channels initiates an RWP. le) 480 VAC ESSENTIAL BUS UNDERV0LTAGE PROTECTION lPlantElectricalSystem-Loss Each 480 VAC essential bus has three undervoltage relays set at

      >278 volts (58% of 480 volt nominal), arranged in two-out-of-three Togic per bus and connected to a 30 + 1.5 second time delay. Should power not be restored on two of the tEree 480 VAC essential buses before the 30 second time delay, a reactor scram will be initiated.

l Degraded Voltage and Loss of Voltage Protection under degraded conditions is provided for the essential power system by the use of undervoltage relays to produce appropriate action corresponding to the degree of voltage degradation. Each 480 VAC essential bus has three undervoltage relays set at 416 + 20 volts (86.7% of 480 volt nominal) arranged in two-out-of-threi logic, one relay per phase. These relays are individually

Fort St. Vrain #1 Technical Specifications Amendment No. Page 4.4-13b O aiarmed, end connected to 120 + s second timers. In the event of a degraded power situation where two-of-the-three relays trip and remain tripped for the 120 second time period, the main power circuit breaker for that bus is opened. Timing out of the 120 second time delay which causes the affected 480 VAC essential bus to be deenergized, or voltage dropping below 372 + 11 volts (77.5% of 480 volt nominal) in less than 120 seconds, will~ actuate inverse time delay undervoltage relays arranged in two-out-of-three logic per bus. Actuation of these inverse time delay relays will attempt to restore power to the affected bus by Automatic Throwover (ATO) to its neighboring 480 VAC essential bus. Interlocks are provided to prevent connecting more than two 480 VAC essential buses together. On a loss of 480 VAC bus voltage caused by the loss of all outside power, a second set of inverse time delay undervoltage relays set at 328 + 10 volts (68.3% of 480 volts nominal) and arranged in two-out-of-tWree logic per bus will be tripped. Loss of voltage on two-out-of-three 480 VAC essential buses or tripping of two-out-of-three main power circuit breakers caused by the degraded voltage 120 second time delay will; 1) trip all three main power circuit breakers, 2) start the diesel generators, 3) load shed on all three buses, 4) close both diesel generator breakers and sequence the loads onto the diesel generators. The tie is established or re-established to the 480 VAC essential bus which is first energized by the diesel generator. Essential Bus 2 is interlocked so it can b3 connected to only one of the other two 480 VAC essential buses. O i 8esis for Actions The required actions and the time allowed for those actions has been selected to incorporate the intent of NRC's Standard Technical Specifications for LWR's, while taking into account the unique design features of the FSV High Temperature Gas Cooled Reactor. Time (12 hours, 24 hours or 48 hours depending on the function) has been allowed immediately after discovery of an inoperable channel to permit repair without tripping the channel since the liklihood of a plant transient coupled with another channel failure during this time period is small. Placing a channel in the tripped condition may cause an unnecessary transient and has been avoided as much as possible. The reduction of manual trips while other protection functions are available will assure plant safety while minimizing challenges to plant safety systems and unnecessary transients. Additional time (12 hours) has been allowed to. permit the placing l of the plant in a MODE where the limitation does not apply if tripping the inoperable channel would cause a transient. This additional time will allow the plant to utilize normal procedures to achieve the new MODE. When complete SHUTDOWN is required 12 hours has been allowed to The ! place the plant in STARTUP and 12 hours more to reach SHUTDOWN. ' o first 12 hours is chosen to be consistent with reaching a MODE where V the limit does not apply and the second 12 hours is allowed to l

i Fort St. Vrain #1 Tecnnical Specifications i J Amendment No. Page 4.4-13c ,

                                                                                      )

O i icomPeteanorderiySauro0WN. l Drift Operation with a trip set less conservative than its TRIP SETPOINT but within its specified ALLOWABLE VALUE is acceptable on the basis that the difference between each TRIP SETPOINT and the ALLOWABLE VALUE is equal to or less than the drift allowance for all trips including those trips assumed in the safety analyses. Where the drift is only measured by a CHANNEL CALIBRATION no margin is shown between the TRIP SETPOINT and ALLOWABLE VALUE. O O

O P-85214 ATTACHMENT 3 Significant Hazards Considerations Analysis O

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Attachment 3 P-85214 SIGNIFICANT HAZARDS CONSIDERATIONS ANALYSIS I. INTRODUCTION This Significant Hazards Considerations Analysis describes the reevaluation of Fort St. Vrain (FSV) Nuclear Generating Station Plant Protective System (PPS) setpoints, and describes the method used to determine those setpoints. The results of this setpoint reevaluation require a revision to the Facility Operating License, Technical Specification LSSS 3.3, Table 3.3-1, LC0 4.4.1, and Tables 4.4-1 through 4.4-5. In response to a Comission letter from Speis to Fuller, dated August 28, 1978, PSC reevaluated FSV PPS setpoints. As a result, PSC informed the Commission, in a letter from Lee to Miller dated October 1, 1980 (P-80340), that the first phase of identifying Instrument Trip Settings and Absolute Values for Category I and II setpoints contained in LSSS 3.3, Table 3.3-1 and LC0 4.4.1, Tables 4.4-1 through 4.4-4 had been completed. In this submittal, the existing Technical Specification Trip Setpoint was redefined as the Absolute Value, and a new Trip Setpoint was calculated. The minimum difference between the Absolute Value and the new Trip Setpoint was the instrument channel cumulative inaccuracy, determined by the least squares method utilizing manufacturers' published accuracy data. Included in PSC's October 1980 submittal, was supportive information for the proposed new Absolute Values.for Low Circulator Speed, Fixed r Feedwater Flow - Low, Loss of Circulator Bearing Water and the applicability of 5% and 30% Rod Withdrawal Prohibits. In a Comission letter from Madsen to Lee, dated July 11, 1983 (G-83255) PSC was informed that their October 1980 submittal did not adhere to the current practice used for the Standard Technical Specifications (STS). PSC utilized Absolute Values rather than Allowable Values in their proposed Technical Specifications. Setpoints in the STS are defined as limits with either greater than or less than, in contrast to the tolerances with plus or minus used by PSC. In addition, PSC defined a reportable occurrence as exceeding an Absolute Value, as opposed to an Allowable Value. As a result, in their letter The Commission recomended that FSV PPS setpoints be specified in terms of an Allowable Value and a Trip l Setpoint, " expressed as either greater than or less than as well as I equal to the value specified." The Commission also informed PSC I that the method described by the Commission was consistent with industry consensus as stated in ISA Standard S67.04-1982, "Setpoints for Nuclear Safety-Related Instrumentation used in Nuclear Power. Plants" and that the standard would be endorsed by a Commission Regulatory Guide in the near future. (A draft of ISA Standard S67.04 was endorsed by a draft of proposed Revision 2 to Regulatory Guide 1.105, issued for comment in December 1981. Subsequently, this issue I of ISA-S67.04 was finalized in 1982.) The Commission requested in i their letter of July 11, 1983 (G-83255), a discussion on the various types of data used and sample calculations, addressing " test L

Attachment 3 P-85214 g equipment accuracy, process measurement accuracy, environmental v effects, component accuracy and drift rates." It was also recommended that the Technical Specifications for FSV be revised to

   " bring them more in line with the fonnat and degree of specificity used in STS for water reactors."

In order to clarify the method used in the calculation of setpoints, a meeting between PSC and the Comission was held on October 27, 1983. In this meeting, the application of ISA Standard S67.04-1982 to FSV PPS Setpoints was discussed. As described in a Comission letter dated November 3, 1983 (G-83409), PSC agreed to revise the 1980 application to include the STS type format, the analysis of Category III instruments, as defined in P-80340, and bases for the numerical values selected. In a letter dated March 9, 1984 (P-84078), PSC enclosed the work specification for the setpoint reevaluation program for the Commission's review, which incorporated the method recomended by the Comission. In addition, PSC agreed to resubmit the proposed amendment to the Facility Operating License. This Significant Hazards Considerations Analysis supports the proposed amendment. a ( Attachment 3 P-85214 II. METHODOLOGY The method used by PSC for the reevaluation of PPS setpoints was requested by the Comission in the October 27, 1983 meeting between PSC and the Comission and submitted to the Commission in'a letter dated March 9, 1984 (P-84078). ISA Standard S67.04-1982 was used by PSC as a guideline, and the standard, in draft form, was endorsed in the draft of proposed Revision 2 to Regulatory Guide 1.105, issued for comment in December 1981. Therefore, the method used as guidance for reevaluation has been recomended for use by both the nuclear industry and the Commission. A schematic of PSC's adaptation of the ISA methodology has been included as Figure 1. ISA Standard S67.04-1982 defines Safety Limits as " limits upon important process variables which are found to be necessary to reasonably protect the integrity of certain physical barriers which guard against the uncontrolled release of radioactivity." Specifically, in the FSV Technical Specifications, " Safety Limits are defined to protect the fuel particle integrity and the integrity of the primary reactor coolant system boundaries. The integrity of these barriers will ensure that an uncontrolled release of radioactivity could not occur." Thus, the specific physical barriers protected at FSV are the fuel particle coating and the primary coolant system boundaries. The FSV Technical Specifications also define the Limiting Safety System Settings (LSSS), which "are established for instrumentation Q and protective devices related to the process variables upon which Safety Limits are based." The LSSS parameters monitored by instrumentation at FSV are listed in Technical Specification LSSS 3.3, Table 3.3-1, as follows: High Neutron Flux, High Reheat Steam Temperature, Low -Primary Coolant Pressure. High Primary Coolant Pressure, and High Moisture in the Primary Coolant. ISA S67.04-1982 applies only to LSSS parameters associated with Safety Limits. In the October 27, 1983 meeting between PSC and the Commission, the Comission requested that PSC apply the ISA Standard to all FSV PPS parameters. The majority of FSV PPS parameters protect safety related equipment and are not associated with the Safety Limits. Therefore, the Comission requested PSC to identify an " Analysis Value" for each parameter, whether it be an LSSS parameter or not. The Analysis Value is the value of a parameter for which a trip and initiation of automatic protective action is assumed to occur in FSV accident analyses. Provided that the trip occurs at a value equal to or more conservative than the Analysis Value, analyses demonstrate that consequences of the accident or transient are acceptable and do not exceed Safety Limits or equipment design limits. Since safety analyses documented in the FSV FSAR assume that trip and initiation of automatic protective action occur at the least conservative trip setting permitted by the current Technical 5 L

Attachment 3 P-85214 Specifications Tables 4.4-1, 4.4-2, and 4.4-3, for any parameter, (-) these Trip Settings are the Analysis Values. The Comission requested that PSC apply the ISA Standard methodology to these Analysis Values to arrive at Allowable Values and Trip Setpoints for each PPS parameter. The factors which are identified by thL ISA Standard and considered in the determination of Analysis Values are: (1) The effects of potential transient overshoot. (Section 4.3.1 (4)) (2) The effects of transient time response characteristics. (Section 4.3.1 (5)) (3) Process measurement inaccuracy. (Section4.3.1(3)) The effects of these factors are discussed for each Analysis Value in Attachment 4. The ISA Standard states that "For each LSSS a Trip Setpoint and its associated Allowable Value shall be established." An Allowable Value is defined in the ISA Standard by the allowances for instrument error between the Allowable Value and the Safety Limit. These allowances are divided into six factors in Section 4.3.1 of ISA S67.04-1982. Three of these factors are accounted for in the r] determination of the Analysis Value (as described above), using PSC's V revised ISA methodology. The remaining three factors contributing to instrument error and used to determine the Allowable Values have been addressed in the following manner: (1) Accuracy (including drift) of components not tested when the setpoint is measured. (Section 4.3.1 (1)) This factor was determined by PSC as the accuracy, including drift, of components not tested when the monthly / quarterly surveillance is performed. The greater of two values, either the manufacturer's inaccuracy or the yearly drift, was used in the Allowable Value calculation. (2) Accuracy of test equipment. (Section4.3.1(2)) Test equipment accuracy was determined by PSC using the manufacturer's quoted accuracy of the least accurate equipment permitted to be used for a given surveillance. (3) Environmental effects on equipment accuracy (Section 4.3.1 (6)) These were determined using the environmental qualification report and manufacturer's data. The use of the environmental qualification report and method of application 0, Attachment 3 P-85214 in many cases resulted in the use of a more conservative fl value than required by the ISA Standard. The environmental V qualification report gives the most extreme conditions for which the equipment is qualified. The ISA Standard requires only that the conditions which are a result of the accident which that particular component is required to mitigate be considered. Also, the effects of seismic and environmental events were combined as if these events occurred simultaneously, which is more restrictive than the design basis accident analysis for FSV. The items considered above were combined using the square root of the sum of the squares where. no common source error existed and algebraically when the effects could occur concurrently. A " total inaccuracy" value was calculated which was used to determine the margin between the Analysis Value and the Allowable Value. The Trip Setpoint, according to the ISA Standard, is to be established ~by determining the margin for drift between the Allowable Value and Trip Setpoint. This margin is defined by the ISA Standard as " Drift of that portion of the instrument channel which is tested when the setpoint is determined." In the October 27, 1983 meeting, the Commission further specified that the test of the instrument channel to be utilized for this drift consideration is the monthly functional test, as opposed to the annual calibration test. The drift of the latter is taken into consideration in the allowances between the Analysis Value and the Allowable Value. For certain r] V parameters, the portion of the instrument channel which is tested monthly is checked only for logic operability by pulse testing. Therefore, the allowable value and the trip setpoint are the same for those parameters. Two years of monthly surveillance data was reviewed to determine instrument channel drift. In the process of applying ISA Standard S67.04-1982 to FSV PPS setpoints, certain Trip Setpoints resulted which could infringe on the normal operating range of their parameters. Operation with these setpoints would result in unnecessary reactor scrams, loop shutdowns i and circulator trips occurring during minor transients under normal operating conditions. In these cases, an evaluation was performed to t determine if a less conservative Analysis Value could be used, which ! continued to protect the Safety Limit or equipment design limit. The ! reanalyzed Analysis Values and their resulting Trip Setpoints I continue to provide overall plant safety. Their reanalysis will reduce the number of challenges to plant systems without a significant reduction in margins from the Safety Limit or equipment design limit. The reanalyzed Analysis Values are included in Attachment 4, and discussed in detail in the following section. O Attachment 3 P-85214 The Rod Withdrawal Prohibits, Table 4.4-4 of the Technical S Specifications, were not analyzed in this program. They are ("'/

  \_          administrative in nature and no credit is taken for them in accident analyses; therefore, the ISA Standard does not apply. Trip Setpoints and Allowable . Values for those parameters have been established to assure operator compliance with Administrative Procedures to maintain the Interlock Sequence Switch in the proper position consistent with the reactor power level.

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• Attachment 3 t P-85214

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III. EVALUATION . A. PRIMARY C00'LANT PRESSURE - LOW Primary Coola'nt Pressure - Low is listed in FSV Technical Specification LSSS 3.3,cTable 3.3-1. Althoughi low primary coolant pressure itself is ntat a Safety Limit, it is an indication of inadequate core cooling and protects PC3V internal thermal temperature limits. The present Technical Specification Trip Setting of i 50 psi below normal gage pressure, programed with circulator inlet temperature, .w as established for depressurization accidents analyzed in FSAR sections 14.11, 4.3.3, 14.7 and 14.8. JNo. " Rapid Design Basis Accident 2 (DBA-2) Depressurization/ Blowdown," FSTR Section 14.11, assumes ' primary coolant pressure decay is instantaneous in the analysis of subsequent cooling.'Therefore, the consequences of DBA-2 are not affected by changing the Primary Coolant Pressure - Low Analysis Value. For the Maximum Credible Accident analyzed in FSAR Section 4.3.3 and 14.8, multiple failures are assumed in conjunc. ion with the offset rupture of 2 inch diameter helium purification system regereration piping. The accident is assumed to occur from 100% rated ' power. The PCRV pressure decays due to a loss of primary coolant inventory at a rate corresponding to a time constant of about 1600 seconds. Primary coolant pressure drops 50 psi below normal at about 120 seconds, at which time primary coolant flow is 97% of bf s rated, and the average core outlet temperature has peaked at 13 degrees F above normal for 100% power operation. For an analysis value of 90 psi below normal, reactor scram occurs at about 220 seconds after the loss of helium is initiated. This results in primary coolant flow of 92.5% of rated flow at the time of the trip, and a peak average core outlet ~ temperature of 44 degree F above normal for 100% power operation. After the reactor scram, the core outlet temperature declines steadily with continued core cooling by the continually decreasing primary coolant inventory. The -effects of an ingress of air due to the decreasing primary coolant temperature after completion of the depressurization are analyzed in Section 14.11.2.3.1 of .the FSAR. This analysis hypothesizes are instantaneous primary coolant system depressurization. As a result, reactor scram is also instantaneous and therefore the low primary coolant pressure setpoint does not affect the consequences of the accident. Analyses were performed to determine the effects of an Analysis Value of 90_psig below normal by evaluating the end points at 100% and 25% load using the SCAP heat balance code. The results

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demonstrate the operation at 90 psi below normal is acceptable. The reduction in helium pressure from 700 psia to 610 psia at 100% power, and aboct 615 psia to about 525 psia at 25% power, results in a 0 _ _ _ _ _ , _ _ _ __

Attachment 3 P-85214 reduced helium density. The circulator speeds increase from about

 .O 8263 rP= to about 9379 rP= at too5 Power. aad from about 2583 to about 3004 rpm at 25% power, to compensate for the reduced helium density. Because the circulator speeds increase, helium flow does not drop significantly.          Helium flow decreases to about 3,484,285 lb/hr from 3,532,213 lb/hr at 100% power, and decreases to about 1,053,262 lb/hr from about 1,059,439 lb/hr at 25% power. Core power to flow ratio changes only .01 at both 25% and 100% power. Helium temperatures at the steam generator inlet only change from about 1383 degrees F to about 1392 degrees F at 100% power, and from about 1192 degrees F to about 1194 degrees F at 25% power.

As demonstrated by the analyses described above, the effects of changing the Analysis Value for Primary Coolant Pressure - Low from 50 psi to 90 psi below normal programed pressure are not significant. The new Analysis Value. continues to protect against inadequate core cooling and exceeding PCRV internal temperature limits. B. PRIMARY COOLANT PRESSURE - HIGH High primary coolant pressure indicates continued steam / water leakage into the PCRV and serves as a backup trip to the PPS moisture monitors. For significant steam leakage into the PCRV, compounded by coincident failure of the redundant moisture monitors, the high primary coolant pressure trip initiates reactor scram, shutdown and p dump of a preselected loop, and main steam depressurization of the v remaining operating loop. These actions are designed to prevent opening the PCRV relief valves. The high primary coolant pressure trip is programmed with reactor power (using circulator inlet temperature), with the most severe consequences, in terms of graphite oxidation, occurring at 100% power. The existing value of 7.5% above normal, programmed with reactor power, has been reanalyzed to justify an Analysis Value of 70 psi above normal gage pressure. The Analysis Value and additional supporting information have been included in Attachment 4. FSAR Section 14.5.3 analyzes six accident cases resulting from steam ingress into the PCRV. Cases 1 and 3 are not included here because they analyze events which are detected by the moisture j monitors, so the high primary coolant system pressure setpoint is not reached. Cases 2, 4, 5 and 6 postulate multiple failures'in the moisture monitor system, resulting in a high primary coolant pressure trip. ~Therefore, these cases have been reanalyzed for a high primary coolant pressure Analysis Value of 70 psi above normal, programmed with reactor power by utilizing circulator inlet temperature. Since a leak from'100% power causes the most severe consequences in terms of graphite oxidation, only a trip at 100% power has been reanalyzed. Multiple failures must take place in the moisture monitor system before action by the high primary coolant pressure trip occurs, O V t

Attachment 3 P-85214 therefore no additional failures in the high primary coolant pressure system are assumed in the accident scenarios. {') Fo.r each case reanalyzed, the same conditions are applied. Cooling is continued with the remaining operating main loop for about 30 minutes, at which time the temperature of the core and the steam generators is about 500 degree F. If, at this time, PCRV pressure indicates significant leakage in the operating loop, the economizer-evaporater-superheater sections are isolated and dumped, and the reheater sections in that loop are flooded. No credit is taken in these reanalyses for moisture removal by condensation in the primary circuit excluding the graphite, or by the helium purification system. FSAR 14.5.3.2 Case 2 - Subheader Rupture and Wrong Loop Dump PCRV Pressure, steam content and maximum graphite temperature during this accident scenario are shown in Figure 2. As a result of the failure of the redundant moisture monitors in the leaking loop, the high moisture level is detected by the monitors in the intact loop in about 13.6 seconds. The intact loop is then tripped, shutdown and dumped, concurrent with a reactor scram. The leaking loop, continuing to operate, leaks at the equilibrium rate of 21.5 lb/sec. Cooling after the scram tends to reduce the system pressure, but the continued steam / water leakage and steam formation will cause the pressure to rise. In about 2 minutes the Analysis p Value of 70 psi above normal pressure, programed with circulator (_ inlet temperature, is reached due to the reduction in circulator inlet temperature. At this time, a feedwater pressure reduction is initiated for the leaking loop which further reduces the leakage rate. A maximum . primary coolant pressure of 739 psia is reached approximately 10 minutes after the start of the leak. For the next 20 minutes the PCRV pressure rises and falls, reaching a value of 648 psia at the end of the 30 minute main loop cooldown. From this point on, cooling is accomplished with the flooded reheaters, and the main 'section of the leaking steam generator is isolated and dumped. During the cooldown utilizing the flooded reheaters, the pressure rises to 658 psia but the steam-graphite reaction is negligible due to the low temperature of the graphite. The steam / water leakage for this case is 15,000 lbs. of which only 180 lbs. react with the core graphite as shown in Table 1. FSAR 14.5.3.4 Case 4 - Subheader Rupture with Moisture Monitor Failure and Correct I.oop Dump For this accident, the redundant moisture monitors in both loops are assumed to fail to detect and dump the leaking steam generator. Water / steam continues to leak into the primary coolant system until the Analysis Value, equivalent to 775 psia for this case, is reached at 157 seconds. A reactor scram is initiated and the correct loop is isolated and dumped. Of the 3200 lbs. total H2O inleakage,1112 lbs. v

l Attachment 3 P-85214 g react with the core graphite, as shown by Figure 3. In this case, , y reactor cooling utilizing feedwater is maintained by the remaining I operating loop for the total duration of the accident since the I leaking loop was correctly isolated and dumped. FSAR 14.5.3.4 Case 5 - Subheader Rupture with Moisture Monitor Failure and Wrong Loop Dump In this case, the moisture monitors are assumed to fail so that a reactor scram with an immediate turbine trip and a ~ steam generator dump of one loop is initiated on high PCRV pressure (775 psia) at 157 seconds. Further, the wrong loop is dumped but the faulty loop, which is used for the cooldown, is operated at reduced pressure to minimize steam inleakage during the cooldown. The maximum pressure reached during the 30 minutes of main loop cooling is 783 psia at 200 seconds after the start of the leak. During cooling the total H2O inleakage for this case is 15,600 lbs of which 1162 lbs. reacts with the core graphite, as is shown on Figure 4. It is noted that the 15,600 lbs. total H20 inleakage is a slight reduction from the 15,740 lbs. value shown in FSAR Table 14.5-1. The initial evaluation, contrary to Note (3) of the table, assumed for Case No. 5' only, that the leakage was terminated after 30 minutes from scram and not 30 minutes from the start of the leakage. This inconsistency has been corrected in this evaluation. A FSAR 54.5.3.4 Case 6 - Subheader Rupture with Moisture Monitor Failure, Correct Loop Isolation and Failure to Dump In this case, moisture monitors in both loops are postulated to fail to detect the leaking steam generator. Water / steam continues to leak into the primary coolant until the Analysis Value, equivalent to

     .775 psia is reached at 157 seconds. A reactor scram is initiated at this point and the leaking loop is isolated but not dumped. During the 157 seconds prior to isolation of the leaking loop, about 3200 lbs. of H2O will leak into the primary coolant system. Following this initial inleakage, the entire 6000 lbs. inventory of the steam generator       is assumed to enter the primary coolant system.

Conservatively assuming that the isolated leaking steam generator is pressurized to 1000 psia because of steam line isolation valve leakage, the draining rate is specified as the inleakage rate from an operating leaking loop with reduced feedwater pressure (about 8 lb/sec.). The primary coolant pressure reaches a peak of 783 psia 200 seconds after the start of the leak. Figure 5 shows that the total H2O inleakage for this case is about 9200 lbs., of which approximately 1200 lbs. reacts with the core graphite. C. LOW SUPERHEAT HEADER TEMPERATURE AND HIGH DIFFERENTIAL TEMPERATURE BETWEEN LOOP 1 AND LOOP 2 Low superheat header temperature, in conjunction with high differential temperature between loop 1 and loop 2, initiates a loop l l

Attachment 3 P-85214 shutdown. The function of the Low Superheat Header Temperature trip C, is to provide safe shutdown of a loop on early indication of potential superheat'er header floodout. This action will also preclude unnecessary turbine trips when only a loop trip is required to prevent wet steam or water from flowing into the main turbine. The turbine control system, which includes a low main steam temperature turbine trip, is available as a backup to the loop trip. The existing Low Superheat Header Temperature Analysis Value of 800 degree F has been reanalyzed to justify an Analysis Value of 780 degree F. At normal main steam operating pressure of 2400 psig, the saturation (floodout) temperature of the steam is about 660 degree F. The Low Superheat Header Temperature Analysis Value of 780 degree F was selected to provide early indication of a floodout event but to be sufficiently below the expected main steam temperature range to avoid spurious trips during normal plant operation. A 65 degree F High Differential Temperature Between Loops Analysis Value was reviewed to determine if this 15 degree higher value would result in a significant time delay in the initiation of the loop trip after a flow imbalance event begins. Since the operating intact loop main steam temperature will be controlled at the nominal setpoints of about 880 degree F at 30% load and 1000 degree F at 100% load, the loop temperature differences will be between approximately 80 and 200 degree F when the malfunctioning loop reaches the Low Superheat Header Temperature setpoint. n Therefore, High Differential Temperature Between Loops 1 and 2 will U be tripped first, before the Low Superheat Header Temperature trip occurs, even if the trip does ne+ accur until the Analysis Value of 65 degree F is reached. The low main steam temperature setpoint for the turbine trip is 800 degree F, therefore the loop trip interaction with this system was reviewed to determine if trip of a malfunctioning loop occurs first over the normal power operating range of the plant. Main steam from the two loops is mixed upstream of the turbine inlet temperature sensors. The temperature of this mixture is compared with malfunctioning loop temperature at the point at which Analysis Values for low Superheat Header Temperature and High Differential Temperature Between Loops are reached. This analysis is presented here and in Attachment 4 and demonstrates that the conditions for trip of a malfunctioning loop are attained prior to reaching the turbine trip setpoint. 30% Power Malfunctioning Loop Temp: 780 degree F Other Loop Temp: 880 degree F Turbine Mixed Inlet Steam Temp: 830 degree F Loop-to-Loop delta T: 100 degree F

Attachment 3 P-85214 100% Power G J Malfunctioning Loop Temp: 780 degree F Other Loop Temp: 1000 degree F Turbine Mixed Inlet Steam Temp: 890 degree F Loop-to-Loop delta T: 220 degree F From the above it can be seen that conditions for trip of a malfunctioning loop are attained prior to reaching the turbine trip setpoint. Therefore, turbine trip is precluded due to a single malfunctioning loop over the plant power operating range. The new Analysis Values for Low Superheat Header Temperature and High Differential Temperature continue to ensure that steam generator floodout does not occur. In addition, they ensure that unnecessary trips of the turbine do not occur due to low main steam temperature. D. CIRCULATOR SPEED LOW Low circulator speed initiates a circulator trip to avert a loop shutdown as a result of only one malfunctioning circulator. Low circulator speed indicates speed control or equipment malfunction, resulting in decreased helium flow. This may cause a mismatch between heat input and heat removal (by feedwater flow) in a steam generator, resulting in a loop shutdown due to Low Superheat Header Temperature - High Differential Between loops (discussed in the o previous paragraphs). The remaining circulator in the affected loop L) is released to cxceed its normal programed speed, allowing operation at up to 50% power on a single loop. The existing Technical Specification trip setting of 1910. rpm below normal as programmed by feedwater flow, has been reanalyzed to justify an Analysis Value of 2390 rpm. Circulator coastdown characteristics are such that circulator malfunctions are detected quickly on the basis of a speed measurement. Upon complete loss of driving power, the time taken to coast down 25%(2390 rpm) from rated speed is 2 seconds; at part load the time would be up to 4 seconds. .PPS action has an intentional delay of 5 seconds to discriminate against transient speed deviations. The time constant of the steam generator superheater header temperature responding to a change in helium flow is approximately 30 seconds. Therefore, a reduction in circulator speed of 2390 rpm (Analysis Value) in less than 30 seconds will result in the trip of a single circulator followed by a power runback (if required) and a speedup of the remaining circulator, thus avoiding a loop trip on Low Superheat Header Temperature. E. FIXED FEEDWATER FLOW-LOW The fixed feedwater flow circulator trip is used to prevent steam generator operation at tube temperatures above design values in the Attachment 3 P-85214 event of a sudden loss of feedwater flow. The Low Feedwater Flow (7 Trip programmed with circulator speed provides protection for reduced U feedwater flow events. There are two aspects to the trip for sudden loss of feedwater flow, the minimum magnitude of feedwater flow and an allowable time delay for trip actuation. A feedwater flow of zero and a time delay of 5 seconds for trip actuation will not result in. the tube metal temperatures exceeding ASME-Code Allowable Values. Three general modes of operation were considered in analyzing plant response to loss of feedwater: two loop operation at 100% and 25% initial load condition and one loop operation at 50% plant load (100% loop load).

1. Sudden loss of Feedwater Flow Analyses were performed to show that allowable generator tube metal temperatures would not be exceeded if feedwater were lost with the plant initially in:

a. Two loop operation at 100% power.

b. Two loop operation at 25% power.

c. One loop operation at 50% power.

Two types of failure were considered. These combinations bracket all credible cases. In the first case sudden feedwater isolation at the feedwater end of the steam generators would, for the time period of interest, permit the inventory to boil off and continue p) t to add to the steam production from the affected loop. Helium circulation consequently remains high for the duration of the accident. In the second case, a feedwater line rupture at the steam generator inlet could, in effect, produce a steam / water dump resulting in termination of steam production from the affected loop in about two seconds. Total steam available to drive the helium circulators quickly drops to 50% and the circulators must accordingly slow down. From the above, the' limiting conditions arise from the first case in which helium circulation remains high for i the accident time, period of intirest. Accordingly, l this case determinfs the required protection for the

steam generator and was selected for analysis. As an upper limit, the helium flow was assumed to remain constant at 100% of rated during the transient.

a. Sudden Loss of Feedwater Flow in a Single Loop l During Two-Loop Operation The system transient analysis was performed using the TAP code for an assumed total loss of O

Attachment 3 P-85214 feedwater flow to a steam generator, starting at 100% and 25% initial plant power. Due to limitations of the TAP code, it was necessary to simulate the sudden loss of feedwater in a three-second ramp from rated plant flow to zero flow. In addition, the TAP code provides only the response of a " nominal" tube within a " nominal" steam generator module. Accordingly, it is necessary to modify the reference data to account for (1) instantaneous loss of feedwater and (2) behavior of the hottest tube in the hottest module. The three second ramp down in feedwater flow was compensated for by analyzing the transient from the time at which flow reached zero. The differences between the behavior of the hottest tube and the nominal tube were accounted for by adding appropriate delta Ts to the nominal tube temperature, and by conservatively assuming that the hottest tube-temperature changes at the same rate as the nominal tube temperature. This assumption is conservative because the temperature differential between the hottest tube and the helium is less than between the nominal tube and the helium, thus the heat transfer rate is less. Using this analysis, allowable delay times for loop shutdown can then be established to preclude tube failure (defined herein as not exceeding the

 - Q-     allowable          tube              temperature   limit    at    the corresponding tube pressure).

The maximum allowable tube temperature is dependent upon the coincident differential pressure to which the tube is subjected. Since the differential pressures to which the tube is subjected are significantly less than design values for the postulated accidents, the allowable temperatures are significantly higher than design. This analysis assumes that the loss of' feedwater results from closure of a feedwater block valve and establishes allowable tube temperatures at an arbitrarily selected time of six seconds after the loss of flow, at which time steam production in the failed loop has fallen to 66% of normal. (This is a conservative approach since the reduction in steaming rate at longer times would result in slightly higher allowable tube temperatures.) Peak tube temperatures, at the time of completing loop shutdown, illustrate that Superheater I outlet is the critical location. The cause of feedwater loss, a steam / water dump due to a line rupture, results in significantly higher allowable temperatures. O l l l l

Attachment 3 P-85214 Once the loss of feedwater flow trip point is O reached and the built-in PPS delay time expires, a signal is generated to shut down the circulators in the affected loop. Circulator speed valve stroking time, full _ open to close, is 1.5 seconds. After closure of the speed valves, the affected circulators will coast down quickly due to the bearing and gas retarding torque. With the circulators in the other loop still operating and providing a loop flow resistance, the shutdown circulators will go into stall early in their coast down time. At this point, the shutoff valves will close, thereby stopping all_ helium flow through their loop. The coast down time from the speed valve closure to stall has been determined to be 0.3 and 0.7 seconds at 100% and 25% plant power, respectively. The allowable PPS delay after reaching the proposed minimum 0% trip setting to reach the design allowable metal temperature for the hottest tube exceeds the five-second PPS delay currently incorporated into this PPS Analysis Value. The margins are much larger at the reduced initial 25%' plant load conditions.

b. Sudden Loss of Feedwater Flow in a Single Loop O During Single Loop Operation It is assumed that the accident is initiated from a condition where only one loop is operating and i the plant is producing about 50% of rated power.

Thus, the operating loop is running at about 100%

of loop design power. The major difference between this case and that previously considered is that there is no additional source of steam available to maintain helium circulator speed and thus helium circulation decays rapidly with onset of the accident if the accident is due to blockage of feedwater at the steam generator inlet. If due to a line rupture causing a steam / water dump, the almost instantaneous depressurization of the steam

! generator would cause an even more rapid loss of helium circulation. A second difference between this and the previous i case is that there is no restraining torque on the helium circulators in the failed loop due to continued operation of the circulators in the l second loop. Thus, the circulator coast down time is longer and will continue down to about 500 rpm, at which time the helium shut-off valve closes O L

Attachment 3 P-85214 under its own weight. The time for this to occur Q is 12 seconds. This sudden loss of feedwater flow does not permit retention of-the constant- helium flow' and thus limits the metal tube temperature rise. The same corrections are made for the hottest tube as opposed to the nominal tube, and for increased allowable tube temperatures based upon the actual tube differential -pressure being less than that for design differential pressure conditions. The times to reach code allowable tube temperature all exceed the PPS cir~cuitry delay time with a trip setting at 0%, therefore an Analysis Value of 5% is acceptable.

2. Steam Generator Boiling Stability

a. Static Stability During the Final Safety Analysis Report review period, boiling stability of the steam generators at low feedwater flow conditions was a major concern because of.possible damage by overheating of the Economizer-Evaporator-Superheater I (EES 1) section tubes. Since then. -two changes have g occurred to significantly lessen this concern.

These are: (1) new experimental data has become available permitting more accurate analysis of boiling instability and (2) plant operating conditions demonstrated by the FSV startup program are different and less conducive to boiling instability tube damage. Static instability, in the extreme, could result in cessation of flow through a few tubes. Should

                 'this occur, the stagnant tubes would quickly approach the temperature of the ambient helium.

Data from DC-1-4 shows that a helium temperature of about 1136 degrees F was anticipated at the inlet to Superheater II, at 25% feedwater flow. Since'this temperature is greater than the maximum allowable temperature of the EES I tubes (1080 degrees F at the operating tube pressure stress), ! plant operation in an area of potential j instability (significantly less than 25% flow) was-precluded by imposition of a low feedwater loop trip set at 20% feedwater flow when the ISS is in the Power position. Plant operation data during the rise-to-power program,-however, has shown that actual operating O I

Attachmsnt 3 P-85214 conditions are significantly different from the original anticipated conditions due to <

 ~O    regenerative heat transfer associated with the steam generators.      (Ref. EPRI NP-760, Project 457-

1) Predicted operating conditions for three cases (26.1%,18.6%,and 15.8% feedwater flow rates) have been analyzed. Analysis of these three cases shows that the steam generators are statistically stable for 26.1% and 18.6% feedwater, but are unstable at 15.8% feedwater. However, potential instability below 18.6% feedwater is now of much less concern for the following reasons:

First, it should be noted that the EES I tubes operate at a significantly lower differential pressure stress (approximately 1800 psi) than design values (e.g., 2920 psi for the economizer lead-in tubes). Thus, the ASME code maximum allowable temperature is considerably higher than that corresponding to the design pressure. For the limiting location, the economizer lead-in tubes, the maximum allowable mean wall temperature is about 952 degrees F compared to the design temperature of 500 degrees F corresponding to the design pressure. Second, helium temperature at Superheater II inlet J decreases with load between 26.1% and 15.8% feedwater flow. Thus, even if unstable boiling conditions are encountered at flow rates below 18.6%, the maximum helium temperature available at Superheater II inlet would be less than 957 degrees F and thus could not result in significantly exceeding the maximum allowable temperature at the limiting tube location (952 degrees F). Note that this approach is quite conservative in that it postulates that a hot gas streak could penetrate the entire EES bundle from top to bottom with no mixing - a highly improbable situation. Any significant gas mixing would, of course, result in much lower gas temperatures and correspondingly less concern about boiling stability.

b. Dynamic Stability Dynamic boiling instability is of concern since it can result in cyclic variation in steam / water flow rates with corresponding cyclic variations in tube wall temperatures. In the extreme, such variations could result in excessive tube bending stresses due to interference between the bundle Attachment 3 P-85214 and the tube support plates (" bear hug") which, in O combination with cyc14c rediei gradient, could lead to fatigue failures.

temperature Three cases were examined for evidence of dynamic instability: 18.6%, 15.8% and 14% feedwater flow rate. Due to computer code limitations, plant operating conditions at 14% feedwater were estimated from the data at the two higher flow rates. The steam generators were shown to be stable down to 14% feedwater flow by use of the dynamic stability code LOOP. Nevertheless, the effect- on the steam generators of dynamic instability was estimated, assuming that the steam generators were actually unstable below about 20% feedwater flow rate. The maximum temperature variation of the tubes, occurring in the evaporator section, would result from fluctuation in the boiling regime from nucleate to film-type, resulting from the cyclic variation in mass flow. Generalized correlations

• show that, for the operating conditions of interest, this could result in a maximum fluctuation of about 36 O degrees F in tube wall temperature. Test data on V both the General Atomic and the French boiling stability test loops indicate that maximum mean wall temperature swings will be on the order of 25

                      - 40 degrees F. There also would be a 20 degree F variation in tube wall radial temperature difference. These conditions would not change the normal operating stress range and the tubes could thus withstand an extremely large number of such cycles without failing due to fatigue.Accordingly, it is concluded that dynamic instabilty of the steam generators at low feedwater flow rates could not cause damage, even if an instability occurs.

A reduction in the setpoint of the Fixed Feedwater Flow - Low PPS trip from the. present Technical Specification Analysis Value of 20% to 5% and

• M. Lasarev, "The Effect of en Oscillating Dryout Point on Evaporator Tube Lifetime," Trans. ANS, 1972 Annual Meeting, June 18-22, 1972, pp. 396.

O I

Attachment 3 P-85214 retention of the present five-second delay does () not jeopardize the operation of the steam generators or create any new accidents in the event of sudden loss of feedwater flow. Time to reach steam generator maximum allowable tube temperatures under the postulated accident conditions all exceed the five-second PPS delay, the circulator speed valve closure, and the time required for stall or coast down of the associated circulators. Unstable flow conditions in the steam generator, should they occur, would not damage the steam generator even if they should persist for long periods of time. However, an alarm will be provided at 20% feedwater flow to alert the operator that the plant is operating in a undesirable region. The incorporation of the proposed change will remove a source of unnecessary loop trips which have occurred during plant transients and thus improve plant reliability. F. LOSS OF CIRCULATOR BEARING WATER Loss of circulator bearing water initiates a circulator trip to ensure sufficient lubrication for circulator bearings. Recirculating water is supplied to each circulator bearing, set at about 170 gpm and at a pressure about 640 psi above primary coolant pressure. Each circulator bearing set includes two journal bearings, a main thrust bearing and a reverse thrust bearing. The recirculating water is normally supplied by 2 of 3 (1 standby) pumps and is referred to as the normal bearing water (NBW). A backup source is available from the feedwater system and is referred to asbackupbearingwater(BUBW). Given the sudden loss of bearing water from both of the above two sources, a third supply is available for safe shutdown of the circulator. This safe shutdown supply consists of a gas pressurizer and water accumulator capable of supplying bearing water for at least 30 seconds at design flow rate. This is adequate for safe shutdown of the affected' circulator. The setpoint at which the accumulator is fired given a sudden loss of bearing water is the subject of this evaluation. The setpoint needs to be sufficiently high to ensure the bearings are not damaged and yet low Attachment 3 P-85214 enough not to cause unnecessary circulator trips during (Q plant transients, including transfers from NBW to BUBW. v

1. Bearing Clearances Normal operating clearances for the circulator bearings are:

Turbine Journal Bearing (centered) 0.0025 in. Compressor Journal Bearing (centered) 0.0035 in. Main Thrust Bearing (centered) 0.0045 in. ReverseThrustBearing(centered) 0.0045 in. A clearance of 0.001 in. is conservatively selected as the minimum clearance to assure adequate lubrication during shutdown of a circulator from 100% speed.

2. Circulator Shutdown Tests Bearing pressure dynamics during circulator shutdown were measured as a part of the RT-368B tests. During these tests, the NBW and BUBW supplies were terminated and an accumulator was fired at the 475 psid setpoint.

Following accumulator firing, momentary dips in bearing cartridge differential pressure were recorded on Brush recorders during the transfer to the accumulator water O SUPPl y. Minimum pressures observed during these dips are: Circulator Minimum Pressure (psid) A 405 B 405 C 375 D 378 For 'C' and 'D' circulators, these dips in delta P occurred within 0.5 seconds after firing of the accumulator and recovered within 1 second to above 400 psid and within 4 seconds to 450 psid. From this data, it follows that if the accumulators were fired at 450 psid instead of 475 psid, the momentary pressure dips would also be 25 psi less, or a minimum of 350 psid for the 'C' circulator.

3. Journal Bearings Each circulator is equipped with two journal bearings.

The purpose of the bearings is to center the shaft in the housing. Load on the journal bearings is solely a 0 Attachment 3 P-85214 l n function of the imbalance on the rotor. Prior to assembly of the FSV circulators, the rotor is balanced t") so that the residual static imbalance and the  : dynamically imposed imbalance due to the coupling moments during rotation is less than 0.2 inch ounces in any plane. This gives a displacement of 0.00004 in, at 10,800 RPM with a bearing water flow generated delta P of 700 psid across the bearing housing. Displacement increases with increasing circulator speed. Since the stiffness of the bearings is generated by the pressure differential across them, as the pressure drops the stiffness decreases' and therefore for the same imbalance load the shaft displacement will increase. Thus, at 350 psid, the shaft displacement will only be 0.00008 in, at 10,800 rpm and imbalance of 0.2 inch ounces. Thus, it can be seen that the minimum delta P that may occur during a sudden loss of bearing water incident has practically no effect'on the journal bearing operating clearances. Further evidence as to the large margin of safety on the journal bearings was demonstrated in tests performed as part of RT-368B. For these tests, the circulators were shut down with pressure drops across the bearings of only 50 psid, generated by accumulator n water flow. This pressure drop is considered adequate U to shut down the circulator from 8000 RPM.

4. Thrust Bearings Each circulator is also equipped with two thrust bearings: a main thrust bearing, and a reverse thrust bearing. Extensive design margin in these thrust bearings occurs at the design operating speed where the generated thrust is -2000 lbs (i.e., the thrust load is

carried by the reverse thrust bearing). The thrust goes through zero at 90% speed and up to a running maximum of +8000 lbs on the main thrust bearing at 30%

speed. The maximum load on the reverse thrust bearing would occur during a rapid PCRV depressurization event while

the circulator is at 100% speed. During this event, the operating reverse thrust load would increase from 2000 lbs to 2900 lbs in conjunction with the helium pressure decay while the reheat steam pressure in the circulator turbine would remain at operating pressure.

For this reverse thrust load, a bearing water pressure of 263 psid is required to maintain a clearance of 0.001 in. on the reverse thrust bearing. O V

Attachment 3 P-85214 The thrust load on the main thrust bearing, as O described above. varies with circuietor speed. eae is about 8000 lbs at 30% of rated speed. This bearing, however, was designed to accept a maximum thrust load of 11,400 lbs. Assuming a minimum 350 psid bearing water pressure as discussed under " Circulator Shutdown Tests", the minimum clearance of 0.001 is maintained with the maximum thrust load (11,400 lbs) at normal circulator operating speeds. The clearance improves with increasing circulator speed as shown in the attached table. Running Clearance for Circulator Speed 350 psid Bearing delta P (RPM) and 11,400 Thrust Load 2,000 0.0010 4,000 0.0011 6,000 0.0012 8,000 0.0014 10,000 0.0018 There is one case resulting from multiple failures, in which the running clearance on the main thrust bearing is reduced .slightly below 0.001 in. This case is an offset steam generator tube rupture with wrong loop dump, producing high PCRV pressure but below the PCRV w relief valves' setpoint. A circulator in the shutdown loop is self-turbining at 300 rpm with atmospheric steam pressure in the downstream piping of the circulator steam turbine. Under the above conditions, the maximum thrust load of 11,400 lbs is experienced. With sudden loss of bearing water, the rotor is stopped by application of the brake within 6 to 10 seconds. Assuming the minimum bearing pressure of 350 psid, the running clearance is reduced during the 6 to 10 second shutdown time to 0.0009 in. This would not cause any damage to the main thrust bearing. In fact, the original (first prototype) circulator tested at Valmont had no brakes and was normally stopped by reducing _the bearing water flow and allowing the thrust runner to rub on the bearing surface. No damage occurred to the thrust runner as a result of this shutdown method. Thus, at any speed above self-turbining, the shaft clearance will be maintained over the conservative 0.001 in. value during a circulator shutdown, protecting the 450 psid Analysis Value. O A_

Attachment 3 P-85214 G. CIRCULATOR SPEED HIGH - STEAM High circulator speed initiates a circulator trip to provide equipment protection for the helium circulators. The limiting accidents leading to circulator overspeed are loss of restraining torque due to blade shedding and reheat steam line ruptures downstream of the circulator turbines. High circulator speed trip settings of 11,000 rpm and 11,500 rpm were analyzed, and resulted in peak speeds of 13,050 rpm and 13,267 rpm, respectively. Extrapolating this data for the 11,700 rpm analysis value results in a peak speed of 13,360 to 13,370 rpm. The design overspeed of the circulators is 13,500 rpm, thus the new analysis value results in acceptable consequences. Steam line ruptures downstream of the circulators were postulated, and an overspeed trip setting of 11,000 was analyzed. The analysis determined that an overspeed of 13,264 rpm would be reached with no control action or trip. This is less than the design overspeed at 13,500 rpm, therefore a trip at 11,700 rpm will not result in circulator speeds beyond design conditions. O O Attachment 3 P-85214 Summary of Applicable Mode Changes Table 4.4-1 Instrument Operating Requirements for Plant Protective System, Scram The S/D Mode is excluded as applicable in the scram inputs because by definition the RMS in "off" already has a scram input in place. la. Manual Scram (Control Room) The S/U, L, P Applicable Modes conform to the model Westinghouse Standard Technical Specifications (WSTS). The WSTS also include the S/D Modes, excluding the R Mode, as applicable if the scram breakers are closed and the control rod drive system capable of rod withdrawal. An equivalent of this provision is proposed to be applicable to the FSV R Mode. Ib. Manual Scram (Outside Control Room) The S/U, L, P Applicable Modes conform to the model Westinghouse Standard Technical Specifications (W STS). The WSTS also include the S/D Modes, excluding the R Mode, as applicable if the scram breakers are closed and the control rod drive system capable of rod withdrawal. An equivalent of this provision is p proposed to be applicable to the FSV R Mode. \ 2a. STARTUP Channel - High The R Mode applicability is consistent with the FSV existing TS. The existing TS state that permissible bypass conditions are those with the RMS in the "run" Mode. 2b. Wide Range Channel Rate of Change - High This Functional Unit is not presently included in the existing TS. Applicable only in the S/U Mode due to positioning of the ISS switch. 3a and 3b. Linear Channels - High (Channels 3, 4, 5 and 6, 7, 8) The S/U, L, P Applicable Modes conforms to the WSTS whereas the FSV existing TS state no permissible bypass conditions exist. The Linear Channels are used to provide accurate Linear Flux measurements in the upper two decades (1.5% to 150%) of power operation, and therefore are not applicable in the R Mode.

4. Primary Coolant Moisture (High Level and Loop Monitor)

The applicability of toe Primary Coolant Moisture Monitors in the P and L MODES is consistent with LC0 4.4.5. The Dewpoint Attachment 3 P-85214 Primary Coolant Moisture Monitors are used to select a loop that h is putting steam / water into the core to prevent releasing Primary Coolant to the Environs due to a pressure build-up in the PCRV, to minimize oxidation of graphite and metal, and to scram the reactor to reduce the heat available for pressure build-up and oxidation. The ~P and L MODES are appropriate because the analytical Moisture Monitors are Operable in the other MODES and Oxidation below 5% power has been evaluated as insignificant and release of Primary Coolant to the Environs due to pressure increase will not occur below 13% power as noted on Page 6 of the Moisture Monitor Injection Test Report submitted to the The Comission May 10, 1984 with P-84138, Lee to Collins, in the Basis for LCO 4.4.5, and in GA report GA-A13677, P-75043, NED-1350, dated October 10, 1975.

5. Reheat Steam Temperature - High The P and L MODES are applicable. In the other MODES there is no steam in the Reheat Steam lines. This measurement is an indicator of an increase in power generation or decrease in steam flow that is not applicable in-the R MODE. In the S/U MODE the Reheat Steam Temperature is well below the 1075 degree F TRIP SETPOINT and protection from a rod withdrawal accident is

        -provided by six Rod Withdrawal Prohibits including the Multiple Rod Pair Withdrawal RWP and three scrams.

6. Primary Coolant Pressure - Programmed Low The P Mode Applicability conforms to the WSTS and is consistent with the existing TS.

7. Primary Coolant Pressure - Programmed High The Applicable Modes are P and L. As documented in page 6 of the Moisture Monitor Response Tests transmitted with P-84138,-

Lee to Collins, dated May 10, 1984; the PCRV relief valve will not lift at less than 13% power operation. The potential PCRV pressure from Moisture Ingress Events vs Reactor Power is illustrated in GA report GA-A13677, P-75043, NED-1350, dated October 10, 1975, figure 4-1. 8, Hot Reheat Header Pressure - Low The P Mode applicability is consistent with the existing TS.

9. Main Steam Pressure - Low The P Mode applicability is consistent with the existing TS.

O

Attachment 3 P-85214

10. Plant Electrical System - Loss The Applicable Modes are S/U, L, P. The existing TS state no permissible bypass conditions exist. In the S/D and R Mode, the control rods are already inserted. There are other undervoltage relays in the protection system to protect equipment for degraded and loss of voltage conditions.

11. Two Loop Trouble Scram Logic The Applicable Modes are P, L, S/U which is consistent with the existing TS.

12. High Reactor Building Temperatures (Pipe Cavity)

The Applicable Modes are S/U, L. P. The existing TS state no permissible bypass conditions exist. This Functional Unit serves as a backup to the Steam Pipe Rupture Detection System and its Trip Setpoint is set to indicate large leaks of lines carrying high energy steam / water which are not applicable in the R Mode. O O

Attachment 3 P-85214 ,q Table 4.4-2 Instrument Operating Requirements for Plant Protective System, Loop Shutdown la-1f. Steam Pipe Rupture, Under PCRV and Pipe Cavity The Applicable Modes are S/U, L, P. The existing TS state no pennissible bypass conditions exist. During S/D and R Mode conditions, decay heat removal is accomplished by operation of one loop (one loop shutdown). For this condition no automatic action of the Steam Pipe Rupture Detection System is required. During the R Mode, the circulators are not driven by steam. 2a-2d. High Pressure, High Temperature - Under PCRV and Pipe Cavity The Applicable Modes are S/U, L P. The existing TS state no permissible bypass conditions exist. During S/D and R Mode conditions, decay heat removal is acccmplished by operation of one loop (one loop shutdown). For this condition no automatic action of the Steam Pipe Rupture Detection System is required. 3a and 3b. Loop 1/ Loop 2 Shutdown Logic The Applicable Modes are S/U, L, P where the Applicable Logic depends on the Mode and, which circulator (s), circulator O drive and loop are required to be operable. The existing TS U states no permissible bypass conditions exist. During S/D and R, decay heat removal is accomplished by operation of one loop (with one loop shutdown). Prohibit circuitry prevents shutting down of the operating loop. 4a and 4b. Circulator 1A(IC) and IB(10) Shutdown - Loop Shutdown Logic The Applicable Modes are S/U, L, P where the Applicable Logic depends on the Mode and, which circulator (s), circulator drive and loop are required to be operable. The existing TS states no permissible bypass conditions exist. During S/D and R, decay heat removal is accomplished by operation of one loop (with one loop shutdown). Prohibit circuitry prevents shutting down of the operating loop. Sa and Sb. Steam Generator Penetration Overpressure - Loop 1/ Loop 2 The Applicable Modes are S/U, L, P where the Applicable Logic depends on the Mode and, which circulator (s), circulator drive and loop are required to be operable. The existing TS states no permissible bypass conditions exist. During S/D and R, decay heat removal is accomplished by operation of one loop (with one loop shutdown). Prohibit circuitry prevents shutting down of (] J

Attachment 3 P-85214 the operating loop. 6a and 6b.' High Reheat Header Activity - Loop 1/ Loop 2 The Applicable Modes are S/U, L P where the Applicable Logic-depends on the Mode and, which circulator (s), circulator drive and loop are required to be operable. The existing TS states no pemissible bypass conditions exist. During S/D and R, decay heat removal is accomplished by operation of one loop (with one loop shutdown). Prohibit circuitry prevents shutting down of the operating loop. 7a and 7b. Low Superheat Header Temperature - Loop 1/ Loop 2 The Applicable P Mode is consistent with the existing TS. 7c. High Differential Temperature Between loop 1 and Loop 2 The Applicable Mode is P.- The existing TS state no permissible bypass conditions exist. Since coincident low superheat steam temperature and a temperature mismatch between loops is required to provide the loop trip because of the ISS switch position and logic, this function is only applicable in the P Mode. , O O Attachment 3 P-85214 Table 4.4-3 " Instrument Operating Requirements for Plant Protective System, Circulator Trip

1. Manual Trip (Steam)

The All Modes applicability is consistent with the existing TS. The manual trip has been separated to identify both steam and water trips. The applicability depends on the circulator operability requirements of LC0 4.2.1 and 4.2.2.

2. Circulator Speed - High (Steam)

The All Modes applicability is consistent with the existing TS except that a provision is applied depending on circulator operability requirements of LC0 4.2.1 and 4.2.2.

3. Circulator Drain Malfunction The Applicable Modes are R, S/U, L, P and S/D when driven by steam which are consistent with the existing TS.

4. Manual Trip (Water)

The All Modes applicability is consistent with the existing p TS. The manual trip has been separated to identify both steam d and water trips. The applicability depends on the circulator operability requirements of LC0 4.2.1 and 4.2.2.

5. Circulator Speed - High (Water)

The All Modes applicability is consistent with the existing TS except that a provision is applied depending on circulator operability requirements of LC0 4.2.1 and 4.2.2.

6. Circulator Speed - Low Programmed The Applicable Mode is P which is consistent with the existing TS.

7a and 7b. Loop 1/ Loop 2 Fixed Feedwater Flow - Low The Applicable Mode is P which is consistent with the existing TS. 8a and 8b., Loop 1/ Loop 2 Programmed Feedwater Flow - Low 9a and 9b. (one and Two circulators) This Functional Unit is not included in the existing TS. It has been added and is applicable in the P Mode. O

I Attachment 3 P-85214 10a and 10b. Circulator Seal Malfunction - High/ Low O The All Modes applicability with noted permissible bypass conditions is consistent with the existing TS.

11. Loss of Circulator Bearing Water The All Modes applicability is consistent with the existing TS except that a provision is applied depending on circulator operability requirements of LC0 4.2.1 and 4.2.2.

12. Circulator Penetration Overpressure The All Modes applicability is consistent with the existing TS except that a provision is applied depending on circulator operability requirements of LC0 4.2.1 and 4.2.2.

O O Attachm:nt 3 P-85214 Table 4.4-4 (q

.J Instrument Operating Requirements for Plant Protective System, Rod Withdrawal Prohibit

1. STARTUP Channel - Low Count Rate (Channels 1 and 2)

The Applicable Modes are S/U and R which are consistent with the existing TS. This Functional Unit may be bypassed above E-03% power. 2a and 2b. Linear Channels - 5% RWP The Applicable Modes is S/U. The existing TS allows for bypass conditions if the bypass also causes single channel scram. This Functional Unit is an operating indication of the proper position of the ISS switch. 3a and 3b. Linear Channels - 30% RWP The Applicable Mode is L. The existing TS states no permissible bypass conditions exist. This Functional Unit is an operating indication of the proper position of the ISS. 4a. Wide Range Channel Rate-of-Change - High O v The Applicable Mode is S/U. The existing TS do not include this Functional Unit. 4b. STARTUP Channels Rate of Change - High The Applicable Mode is S/U but may be bypassed above E-03% power. The existing TS do not include this Functional Unit. Sa and 5b. Linear Channels - High Power RWP The Applicable Modes are S/U, L. P. The existing TS do not include this Functional Unit.

6. Multiple Rod Pair Withdrawal This Functional Unit is not included in the existing TS. It has been added and is applicable is in All Modes with a note applying to the R Mode.

O Attachment 3 P-85214 Table 4.4-5 Instruinent Operating Requirements for 480 VAC Essential Bus Undervoltage Protection

1. Plant Electrical System - Loss See Table 4.4-1

2. Degraded Voltage This Functional Unit is not included in the present TS. The Applicability is All Modes.

3. Loss of Voltage - Automatic Throwover (ATO)

This Functional Unit is not included in the present TS. The Applicability is All Modes.

4. Loss of Voltage - D.G. Start, Load Shed and Load Sequence This Functional Unit is not included in the present TS. The Applicability is All Modes.

O O i

7, , ( x V (/ TABLE 2

SUMMARY

OF STEAM LEAK ACCIDENT CASES REANALYZED BASED ON Tile COMPLETE OFFSET RUPTURE OF A STEAM GENERATOR SUBi!EADER No. of Scram Peak PCRV Total 1120 Total 1120 Total Graphite Independent Time Pressure Inleakage Reacted Fraction Case Number and-Conditions Failures (sec) (psia) (1b) (ib) Reacted (1) 13.6 739(4) 15,000 180 2.08 x 10-4

2. t!rong Loop Dump (scram, but isolation and dump 1 of intact steam generator at 13.6 sec) + MMS (2) 157 775 3,200 1112 1.28 x 10-3

4. Moisture Monitors Fail (scram, correct loop 1 isolation and steam generator dump at 775 psia) + MMS (2)

Moisture Monitors Fail and Idrong Loop Dump 1 157 783(4) 15,600(3) 1162 1.34 x 10-3 5. (scram, but isolation and dump of intact steam + MMS (2) generator at 775 psia) - Moisture Monitors Fail and Failure to Dump 2 157 783 9,200 1195 1.38 x 10-3 6. (scram, correct loop isolation, but failure + MMS (2) to dump the leaking steam generator at 775 psia) (1) fraction of active core and top and bottom reflector graphite (~600,000 lb) (2) Moisture Monitor System--two independent faults would have to occur to prevent the moisture monitors functioning (3) Leakage terminated at 30 minutes (4) Peak Pressure during first 30 minutes of transient. For Case 2 the pressure rises to 658 psia during cooling on the rehester. For Case-5 the pressure rises to 679 psia during cooling on the reheater.

Attachment 3 P-85214 O SAFETY LIMIT OR a EQUIPMENT DESIGN LIMIT TRANSIENT OVERSHOOT TRANSIENT TIME RESPONSE PROCESS MEASUREMENT ACCURACY

                "                 ANALYSIS VALUE a

ACCURACY UF CCMPONENTS NOT TESTED IN THE MONTHLY SURVEILLANCE ACCURACY OF TEST EQUIPMENT ENVIRONMENTAL EFFECTS ON ECUIPMENT ACCURACY

                 "                 ALLOWABLE VALUE a

DRIFT OF CCMPONENTS O TESTED IN THE MONTHLY SURVEILLANCE

                  'T                TRIP SETPOINT FIGURE 1 SCHEMATIC OF SETPOINT REEVALUATION METH000 LOGY O

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0 l P-85214 ATTACHMENT 4 Analysis Value O

Attachment 4 P-85214 O ANALYSIS VALUE FUNCTION: Reactor Scram FUNCTIONAL UNIT: STARTUP Channel - High ANALYSIS VALUE: E+05 cps PRESENT TECHNICAL SPECIFICATION TRIP SETTING: __

                                                         < E+05 cps LICENSING BASIS: High start-up count rate is provided as a scram for use during fuel loading and preoperational testing or other low power operations.

The trip is only active when the Reactor Mode Switch is in the " Fuel Loading" position (FSAR Sections 7.1.2.3, 7.1.2.8 and Table 7.1-2). The Analysis Value of E+05 cps, which is the same as used at Peach Bottom, would produce a scram on a rod withdrawal accident during low power physics testing conducted in air, with the vessel open, anal {Ref: pJ slightly faster than the 50 seconds assumed in the(The .Commission)ysis. F. E. Swart (PSC) to D. Eisenhut letter dated 7/31/73.) SAFETY LIMIT: None. The specific accident analyzed for Fort St. Vrain (FSV) was a control rod pair withdrawal accident during the initial low power physics testing conducted with the vessel open. Maximum fuel and graphite temperatures were predicted to be less than 800 degree F. This temperature would not result in fuel failure or damage to the primary coolant pressure boundary. Same reference as l given under " Licensing Basis." LIMITING INITIATING EVENT: Contro1 rod withdrawal accident during fuel loading or any in-core maintenance with the Reactor Mode Switch i in the Fuel Loading position. TRANSIENT OVERSHOOT: Not identified in counts per second (cps). l However, for the specific accident analyzed, the power was predicted to peak briefly to about 600MW(t). TRANSIENT TIME RESPONSE: Not identified in the GA analysis. The l scram was assumea to occur at 50 seconds into the rod withdrawal t accident which is somewhat in excess of the time required for automatic scram in high count rate, E+05 cps. lO I i f

Attachment 4 P-85214 O ANALYSIS VALUE PROCESS MEASUREMENT ACCURACY: Not applicable. DISCUSSION: The primary purpose of the STARTUP Channel-High reactor scram was for the initial core loading which was done manually and for the initial low power physics testing which was done with the PCRV open. This scram channel does continue to provide protection for control rod withdrawal accidents during refueling operations or any in-core maintenance operation. The Reactor Mode Switch must be in the Fuel Loading position for this reactor scram trip to be active. O [ Attachment 4 P-85214

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ANALYSIS VALUE FUNCTION: Reactor Scram FUNCTIONAL UNIT: Wide Range Channel - Rate of Change High Channel 3, 4, 5. ANALYSIS VALUE: 5 decades per minute (dpm) PRESENT TECHNICAL SPECIFICATION TRIP SETTING: Currently not contained in the Technical Specifications. LICENSING BASIS: High rate of change of neutron flux rise is used as a scram input during plant start-up and results in additional protection and better scram response than the high-neutron-flux power scram in case of accidental control rod withdrawal. The trip setting is selected to be above the operating rate of flux change of 2 dpm. the Interlock Sequence Q This scram Switch is intrip setting position. STARTUP is active only when FSAR Section 7.1.2.3 and Table (Ref: 7.1-2.) SAFETY LIMIT: None. Assuming the rate of change of neutron flux rise scram fails, plus three other rod withdrawal prohibits also fail and the rod withdrawal accident is terminated by the 140% of rated thermal power scram, no fuel failure is predicted nor is there a breach of the primary coolant boundary (FSAR Section 14.2.2.7). LIMITING INITIATING EVENT: Control rod pair accidental withdrawal at less than 5% power (FSAR Section 14.2.2.7). TRANSIENT OVERSHOOT: None once scram has been initiated. TRANSIENT TIME RESPONSE: The log amplifiers incorporate some time delay in the lower decades in order to minimize false rate-of-change indications. The amplifiers reach 63% of their final output value in 50 msec or less in response to a decade step input within the upper six decades. (FSAR Section 7.3.1.1). See " Discussions" for further details. PROCESS MEASUREMENT ACCURACY: Not applicable.

Attachment 4 P-85214 O ANALYSIS VALUE DISCUSSION: The wide range channel - high rate of change of neutron flux rise is a first line of defense for rod withdrawal accidents at source power and in the startup mode. It is only active when the Interlock Sequence Switch (ISS) is in the "startup" position. The more detailed FSAR analysis in FSAR Section 14.2.2.7 states the rod withdrawal accident is terminated by a 140% power scram 30 seconds into the accident.- In this case, there is not fuel failure nor damage to reactor internals even when primary coolant flow is limited to 10% of rated. However, power does peak to about 2500 MW(t). FSAR Section 14.2.2.7 further states that approximately 15 seconds after accident initiation (rod withdrawal), the power starts rising faster than 5 dpm. A scram or fast reactor period occurs about 6 seconds later and the peak power is reduced s jnificantly compared to the scram at 140% rated thermal power shown in FSAR Figure 14.2.4. The average and maximum fuel temperatures are also considerably lower. O No further amplification of the rod withdrawal accident terminated by a scram on small reactor period is provided in the FSAR. However, additional detail is available from 1969 GA internal correspondence. The 6 second delay is the result of the response time of the rate circuit-period amplifier. Maximum power peaks to 1355 MW(t) occur as a result of the 5 dpm scram as opposed to 2500 MW(t) shown in FSAR Figure 14.2-4 for a 140% power scram. Attached is the predicted rate of flux rise during the rod withdrawal accident terminated by the 140% power scram.

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Attachment 4 P-85214 (J3 y ANALYSIS VALUE FUNCTION: Reactor Scram FUNCTIONAL UNIT: Linear Channel - High Channels 3, 4, 5 Linear Channel - High Channels 6, 7, 8 ANALYSIS VALUE: 140% rated thermal power PRESENT TECHNICAL SPECIFICATION TRIP SETTING: < 140% power 1 LICENSING BASIS: 140% power scram resulted in acceptable accidental consequences (Ref: FSAR Section 14.2.2 Rod Withdrawal Accidents). This same analysis value was also used at Peach Bottom again with acceptable accident consequences. SAFETY LIMIT: At least 390 to 415% of rated power in a rod accident O'withdrawalaccidentasthisistheindicatedpowerwhenthe is tenninated by the backup high reheat steam temperature scram (Ref: GA Document No. 907217 transmitted by GP-2185 dated 2/6/84). In this case less than 5% of the fuel has failed and localized thermal barrier and steam generator insulation damage has occurred. However, the primary coolant pressure boundary is not breached and the accident is judged to be an acceptable one time plant transient (Ref: FSAR Section 14.2.2.6). LIMITING INITIATING EVENT: Accidental rod pair withdrawal from 100% power (Ref: FSAR Section 14.2.2.6) TRANSIENT OVERSHOOT: None in terms of neutron flux once scram action is initiated per GA TAP code. TRANSIENT TIME RESPONSE: <0.2 seconds following a step change in power for the electromatic brakes which hold the control rods in position to be deenergized allowing scram free fall (Ref: FSAR Section7.1.2.3). . PROCESS MEASUREMENT ACCURACY: Control rod shim bank movement can result in decalibration of the ex-core flux monitors. These ex-core flux monitors trip settings are administratively set based upon indicated power at less than 140% of rated power to account for this potential decalibration. Attached Figure 4.14 from GA Document No. O

4 Attachment 4 - P-85214 O . ANALYSIS VALUE 907177A transmitted by GP-2191 dated 2/14/84 are the trip settings based upon indicated power for Cycle 4 operation. The Figure 4.14 trip settings account for decalibration due to control rod shadowing and allowances for other instrumentation errors.

       - DISCUSSION:                 The linear channel high flux monitoring by ex-core monitors provide protection for raactivity insertion accidents of which rod withdrawal accidents are the most limiting (Ref: FSAR

, Section 14.2). The linear channel high trip is backed up by rod withdrawal prohibits, operator manual initiated scram and by the high reheat steam temperature scram. i O-O 6-8

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Attachment 4 P-85214 T. _O ANALYSIS VALUE 1 ( 4 FUNCTION: Primary Coolant Moisture i High Level Monitor Loop Monitor ANALYSIS VALUE: 67 degree F Dewpoint (500 ppm water) for high level monitor

,                                          27 degree F Dewpoint (100 ppm water) for loop monitor PRESENT TECHNICAL SPECIFICATION TRIP SETTING: < 67 degree F Dewpoint for hTgh level monitor
                                                                                < 27 degree F Dewpoint for Toop monitor LICENSING BASIS:                                  The present Technical Specification trip settings were established in FSV License Amendment No. 13, G-76047 dated June i ' O 18, 1976.            The Safety Evaluation Report (SER) attached to this N> license amendment provided the rational for the trip settings. The bases for License Amendment No. 13 were extensive evaluation and insitu testing conducted on the Dewpoint Moisture Monitors (DPMMs) and subsequent- insitu testing of a modified DPMM system. These evaluations and testing were documented in GA-A13677 dated 10/10/75

and GA-A13823 dated 2/17/76.

A primary modification was providing additional insulation and relocating some ' sample lines to assure sufficient sample line temperature to preclude unacceptable condensation of the moisture in the sample line prior to reaching the DPMMs. A second important modification was the replacement of a fixed bypass valve with an . adjustable bypass valve to better assure a sufficient sample flow is provided to the DPMMs along with specifying minimum DPMM flow rates based upon reactor power level. Third, the trip setting was specified in degrees Fahrenheit dewpoint as opposed to parts per million volume water as specified previously. Finally, the temporary exception for the high level monitors which permitted them to be set at a dewpoint corresponding to 5000ppmv during low power operation was replaced with a dewpoint setting of < 67 degree'F (500ppmv). SAFETY LIMIT: None. The DPMMs are a first line of defense for major water ingress events. The backup line of defense is the primary

coolant high pressure scram. Assuming the DPMMs fail to perform the 4

O

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

Attachment 4 P-85214 O ANALYSIS VALUE corrective action or incorrectly perform the corrective action such as dump of the wrong steam generator loop, the backup primary coolant high pressure trip results in no fuel failure or breach of the primary coolant pressure boundary including the lifting of the PCRV relief valves. (FSAR Section 14.5.3). LIMITING INITIATING EVENT: Offset steam generator subheader rupture from 100% of rated power. TRANSIENT OVERSHOOT: Variable dependent upon size of leak in the steam generator tubes or subheaders. The two steam generator tube leaks experienced to date were extremely small and easily terminated by manual operator action before the PPS trip occurred. In a design basis subheader offset rupture accident, the sample delivered to the DPMM will be 3000 to 5000 ppmv water which is equivalent to 120 to 150. degree F dewpoint, far in excess of the trip settings. TRANSIENT TIME RESPONSE: The transient time response is dependent upon a number of variables. The driving force for the primary coolant sample flow is the helium circulator delta pressure which varies with primary coolant helium flow and thus with reactor power level. The capillary sample line to the DPMMs contains a filter which, dependent upon the degree of plugging, will affect the quantity of the sample delivered to the DPMM. Filter plugging can be , and is compensated for by the variable bypass valve position. As filter plugging reduces sample to the DPMM, the variable bypass valve is closed to provide at least the minimum flow as required by Note (t) to Tech Spec LC0 4.4.1. These actions result in an increased transient response time. Finally, the quantity of sample delivered l to the DPMM will affect the fogging time of the mirror. There is an upper limit on sample flow to the DPMMs (75 sec/sec) for the purpose of limiting erosion of the mirror face. The latter is an operational / maintenance concern as opposed to a safety concern. From the above discussion it is apparent that the DPMMs constitute a complex system relative to response time. The desired objective is for.the loop monitors in.the leaking steam generator loop to identify the leaking loop with first-in-with-lockout prior to the steam mixing with the primary coolant and tripping the loop monitors in the opposite or non-leaking loop. The high level monitors, which sample primary coolant from both loops and are required to produce the safety action, must reach their trip setting before the primary O

Attachment 4 P-85214 O v ANALYSIS VALUE coolant pressure high trip setting is reached. This best assures that the leaking steam generator loop.is isolated and dumped, thus limiting the quantity of water / steam ingressed. Finally, the intensity of the light source, which is adjustable, will affect the response time of both the loop and high level DPMMs. The complexity of the systems has long been recognized. The Technical Specification Lc0 4.9.2 required insitu testing with the reactor at three power levels to verify operability of the DPMM system. These tests were run at 5, 25 and 70% of rated power. The results of the tests were reported by P-77144 dated 6/30/77 and by P-84138 dated 5/10/84. The tests did verify the operability of the DPMM system. However, for power levels above 50% of rated, the test data indicated minimum required sample flow rate to the DPMMs of 50 scc /sec as opposed to 40 sec/sec which the tests attempted to verify. 50 scc /sec was and is the minimum DPMM' flow required by Tech Spec LC0

     .4.4.1.

cess MeASuaeMeNT O ea0moisture AccuaAcY: Tnere wili be denietion of the delivered to the DPMM due to condensation in the sample lines. This has been accomodated for by maintaining sample lines within specified temperature limits and establishing the setpoints at sufficiently low levels to assure the depleted moisture level to the DPMM will be sufficient to produce the required safety action (FSV License Amendment No. 13, G-76047 dated 7/18/76). l DIcuSSION: The DPMM system consists of six loop monitors and two ! high level monitors. Three (3) loop monitors each sample primary coolant exiting from the two secondary coolant (steam generator) loops. Their purpose is to identify the leaking secondary coolant loop. No safety action is initiated on trip of the loop monitors alone. The two high level monitors sample primary coolant exiting and common to both secondary coolant loops. In a design basis steam leak accident from 100% power, the expected action would be loop l identification with first-in-with-lockout by the loop monitors at 4.5 i seconds and loop isolation and dump by the high level monitors at 8.6 seconds. (Ref: FSAR Section 7.3.2 Table 7.3-1). These same actions also provide input to two loop trouble which scrams the reactor (Ref: FSAR Fig. 7.1-2 and Fig. 7.1-3). Trip of both high level monitors l without prior loop identification by the low level monitors will result in a reactor scram but no isolation and dump of a secondary coolant loop. Subsequent loop identification by the loop monitors ! l t

l Attachment 4 l P-85214 -O ANALYSIS VALUE would result in isolation and dump of that secondary coolant loop. If none of the above occur or occur improperly, the primary coolant pressure will continue to rise until the high pressure trip setting is reached (157 seconds into the transient). At this time a reactor scram will be initiated, if not already scrammed, and a preselected secondary coolant loop will be isolated and dumped if not already dumped. Additionally, since it is not shown that the correct loop has been dumped in this example, the main steam pressure is reduced to about 800 psig to limit suusequent water ingress into the reactor vessel. All of the above described sequences result in acceptable accident consequences. The core is adequately cooled; limiting graphite oxidation and fuel particle oxidation, and the primary coolant pressure does not reach PCRV relief valve settings (Ref:

                               ~

FSAR Section 14.5.3). O o O

Attachment 4 P-85214 ANALYSIS VALUE FUNCTION: Reactor Scram FUNCTIONAL UNIT: Reheat Steam Temperature - High ANALYSIS VALUE: 1075 degree F PRESENT TECHNICAL SPECIFICATION TRIP SETTING: < 1075 degree F LICENSING ~ BASIS: The reheat steam temperature-high scram is a diverse backup scram trip'to a high-neutronic flux scram in the event of a rod withdrawal accident. The trip levei is chosen to be just above expected transients in order to minimize de scram response time and corresponding temperature overshoots oc rod withdrawal accidents not terminated by a neutronic scram (Ref: FSAR Section 7.1.2.3). For a contro1~ rod withdrawal accident at full power, the RWP at 120% power and the scram at 140%. power ~would have to fail before the diverse reheat steam temperature scram would be required FSAR Section 14.2.2.1 and 14.2.2.6). Q (Ref: SAFETY LIMIT: The rod withdrawal accident terminated by a reheat steam temperature-high scram analyzed in FSAR Section 14.2.2.6 approaches but does not reach a plant safety limit. About 2% of the fuel is predicted to fail. This is less than the average 5% failed fuel which the " Design" gas-borne activities are based (FSAR Section 3.7.4.1.2). The maximum steam generator module inlet temperature is about 2450 degree F which occurs at the module adjacent to the core regions experiencing peak power and temperatures during the excursion. Local hot streaks up to 2600 degree F may occur. The Class B insulation in the steam generator ducts (Hastalloy-x) and the insulation cover sheets in the steam generator would be expected to fail. However, localized failure of these components do not affect either the steam generator performance or the integrity of other reactor intervals. Therefore, fuel failure remains within acceptable limits and there is no breach of the primary coolant pressure boundary. LIMITING INITIATING EVENT: Rod withdrawal accident from 100% reactor power. TRANSIENT OVERSHOOT: 185 degree F per GA analysis which is the basis for the rod withdrawal accident terminated by a reheat steam temperature scram presented in FSAR Section 14.2.2.6. O 1 Attachment 4 P-85214 l O ANALYSIS VALUE TRANSIENT TIME RESPONSE: Variable dependent upon reactor power level per GA TAP Code. For a rod withdrawal accident from 100% of rated power, a delay of 16 seconds occurs due to actual versus indicated steam temperatures (Ref: FSAR Section 14.2.2.1). PROCESS MEASUREMENT ACCURACY: Not applicable. DISCUSSION: The reheat steam temperature-high scram has always included a latch on feedwater flow rate at its existing flow rate at the time of the scram. Feedwater flow is not permitted to runback until reheat steam temperature has decreased to 975 degree F. This is shown as a reset at 975 -degree F on FSAR Figure 7.1-1. The objective was to maintain maximum heat removal until the transient had been arrested as indicated by decreasing reheat steam temperature. A study performed by GA in 1983-84 indicted the latch on feedwater O flow is no longer required. This GA study is GA Document No. 907217 V and was transmitted to PSC by GP-2185 dated 2/6/84. I O

i e, Attachment 4 P-85214 O ANALYSIS VALUE FUNCTION: Reactor Scram FUNCTIONAL UNIT: Primary Coolant Pressure - Low

   . ANALYSIS VALUE: 90 psi below normal, load programmed (Ref: GA Document No. 907790 A dated 1-29-85)

PRESENT TECHNICAL SPECIFICATION TRIP SETTING: < 50 psi below normal, load programmed LICENSING BASIS: Low primary coolant pressure is an indication of helium leakage from the system. A scram is required because the reactor could be inadequately cooled, which would increase the hazard associated with activity release from the PCRV. The trip setpoint is programmed with load to reduce the response time when the plant is at high power. A turbine-generator trip is initiated simultaneously the reactor trip to anticipate-the ensuing decrease in main O.with steam temperature (FSAR Section 7.1.2.3). Based upon analyses, the 90 psi below normal, load programmed analysis value will provide sufficient _ margin for low pressures reached in transient conditions. SAFETY LIMIT: Low primary coolant pressure in itself is not a safety limit for FSV. The concern is adequate core cooling and not exceeding PCRV internal thermal temperature limits. The consequences of depressurization accidents analyzed assuming an analysis value of 50 psi below normal in FSAR Sections 4.3.3 and 14.11 are acceptable. These same depressurization accidents were reevaluated with 90 psi below normal analysis value with little or no change in consequences and thus acceptable (Ref: GA Document No._907790 dated 1-29-85) LIMITING INITIATING EVENT: DBA -2 " Rapid Depressurization/ Blowdown" (F5AR Section 14.11). The accident consequences are not sensitive to the trip setting due to conservative assumptions (Ref: GA Document No. 907790 dated 1-29-85) TRANSIENT OVERSHOOT: Pressure will continue to decrease at a rate consistent with the size of the breach assumed in the primary coolant system and will ultimately equalize with atmospheric pressure.

     -TRANSIENT     TIME RESPONSE:     Na delay assumed in sensing actual pressure. See " Discussion".

O Attachment 4 P-85214

 .O ANALYSIS VALUE J

PROCESS MEASUREMENT ACCURACY: Not applicable. 1 DISCUSSION: Two primary coolant depressurization events have been analyzed for FSV and the analyses for both are presented in the FSAR. DBA -2 " Rapid Depressurization/ Blowdown, FSAR Section 14.11, is not impacted-by the low primary coolant pressure setpoint value because of the very rapid nature of the blowdown. The primary coolant

     . pressure in this case drops 100 psi below nomal within 2 to 3 sec.

The second event is the " Maximum Credible Accident," FSAR Sections 4.3.3 and 14.8, in which multiple failures were assumed in conjunction with the offset rupture of a 2" diameter helium

     . purification system regeneration piping. The accident was assumed to occur from 100% of rated power. Primary coolant pressure drops 50 psi below normal at 120 seconds at which time primary coolant flow is 97% of rated and the average core outlet temperature has peaked to its maximum value of only 13 degree F above normal for 100% power O- operation.       When the.. analysis .value is lowered to 90 psi below normal, the setpoint.is reached 220 seconds into the transient.              In this latter case, the primary coolant flow is 92.5% of rated at the time of the trip and the average core outlet temperature has peaked to a maximum of 44 degree F above normal for 100% power operation.

After the reactor scram, the core outlet temperature steadily declines with continued core cooling and the continually decreasing primary coolant inventory. The re-analysis for the 90 psi below normal analysis value is contained in GA Document No. 907790 A dated 29-85. O Attachment 4 P-85214 ANALYSIS VALUE FUNCTION: Reactor Scram FUNCTIONAL UNIT: Primary Coolant Pressure - High ANALYSIS VALUE: 70 psi above normal rated, load programmed. 775 as shown in psia has been used in the " Limiting Initiating FSAR~ Figure 7.1-14 (Ref: GP-0534 dated 8/2S/80 . ) Event" PRESENT TECHNICAL SPECIFICATION TRIP SETTING: < 7.5% above normal rated, load programmeo LICENSING BASIS: High primary coolant pressure is an indication of steam / water inleakage into the primary coolant system and poses the threat of overpressurizing the PCRV. A scram is required to reduce system temperatures, minimizing steam graphite reaction, particularly as the reactor cooling is immediately reduced by 50% by the shutoff (and dump) of a steam generator loop. In this situation,-the steam O pressure in the second steam generator loop is also reduced (main

 \     stream.depressurization) to limit water leakage to the reactor. (The PCRV relief valve system provides the ultimate protection to prevent the PCRV pressure from exeeding Reference Pressure).            The trip is programmed     with plant load (as indicated by circulator inlet temperature) to reduce the trip time (and corresponding amount of water inleakage) when the reactor is operating at part load. The circulator inlet temperature (essentially the core inlet temperature) corresponds closely with the average bulk gas temperature (most of the reactor helium volume is cold gas) and allows the trip setting to vary with normal primary coolant pressure changes (FSAR Section 7.1.2.3). The analysis value of 10% above normal rated, load programmed will preclude actuation of the PCRV relief valve system while providing adequate margin for plant transients.

SAFETY LIMIT: The PCRV relief valve system provides the ultimate protection to prevent the PCRV pressure from exceeding Reference Pressure (845 psig). However, actuation of the PCRV relief valve system would constitute a temporary breach of the primary coolant

     , pressure boundary.       The PCRV pressure relief system consists of upstream rupture disks and downstream relief valves. The low-set rupture disk has a trip setting of 812 psig ,+1%. Thus, the rupture          ,

disk could open at 804 psig and this pressure would constitute a safety limit regarding breach of the primary coolant pressure boundary. The low-set downstream safety valve has a trip setting of O

Attachment 4 P-85214 q / ANALYSIS VALUE 796 psig +1%. Therefore, the low-set safety valve should open at the lower trii setting of the rupture disk. The low-set safety valve would close at a line pressure of approximately 677 psig; the corresponding primary coolant pressure would be approximately 737 psig (Ref: Tech Spec LSSS 3.3 and FSAR Section 6.8.3). LIMITING INITIATING EVENT: Steam generator feedwater subheader offset rupture from 100% power operation coupled with failure of the dewpoint moisture monitor protective system (FSAR Section 14.5.2 and 14.5.3). TRANSIENT OVERSHOOT: 8 psi above the analysis value of 775 psia (Ref: GP-0534 dated 8/25/80). This provides a margin of 33 psi of not reaching the lower limit of the low-set rupture disk. TRANSIENT TIME RESPONSE: No delay assumed in sensing actual pressure. PROCESS MEASUREMENT ACCURACY: Not applicable. FUNCTIONAL UNIT: Primary Coolant Pressure - High DISCUSSION: The primary coolant pressure - high trip setting is a diverse backup to the DPMM and steam generator dump system for design The FSAR, Section 14.5, analyzes a

                                                            ~

basis steam leak accidents. number of cases where the DPMM and dump system fail to operate, operate incorrectly or do not complete all operations. In these cases the primary coolant high pressure trip setting will be reached resulting in a reactor scram if not already scrammed, dump of a preselected steam generator loop if not already dumped, and finally and in all cases, a runback of steam pressure in the steam generator from about 2400 psig to about 850 psig. This-latter action minimizes subsequent steam inleakage and is instrumental in maintaining primary coolant pressure below the lower.. limit of the PCRV relief valve rupture disk bursting pressure. The proposed change in the analysis value from 7.5 to 10% above normal rated, load programed will extend the time for the high pressure trip setting to be reached from 95 to 157 seconds for the design basis condition of 100% power operation and an offset rupture of a steam generator subheader. This additional margin in the analysis value for the trip setting will provide greater operational i O

Attachment 4 P-85214 0 . ANALYSIS VALUE flexibility.- The margin for not lifting PCRV relief valves is about 33 psi and is adequate. O O

Attachment 4 < P-85214 O . ANALYSIS VALUE 4 FUNCTION: Reactor Scram I FUNCTIONAL UNIT: Hot Reheater Header Pressure - Low-ANALYSIS VALUE: 10 psig (FGLP-1905 dated 7/27/77) PRESENT TECHNICAL SPECIFICATION TRIP SETTING: > 35 psig LICENSING BASIS: Low hot reheat steam pressure is an indication of either a cold reheat steam line or a hot reheat steam line rupture in a section of line common to both loops. Loss of the cold reheat steam line results in loss of the steam supply to the circulators , which necessitates plant shutdown. The direct scram in this case precedes a scram resulting from the two-loop trouble. The loss of either steam line results in loss of plant output, and a reactor 2 scram is appropriate in this situation. The trip point is selected to be below normal operating and transient levels, which vary over a g wide range. (Ref: FSARSection7.1.2.3). The analysis value of 10 psig assures a margin for normal operations and transients while still assuring a fast response in the event of a major reheat steam line rupture. In the event of a complete offset rupture of a reheat header, the maximum rate of blowdown or depressurization occurs in the choke flow regime. The minimum choke pressure for flow of superheated steam to a 12.5 psia atmospheric pressure is-about 10 psig. Below the choke flow regime, the rate of depressurization is significantly slower. The additional delay in detection time for the analysis value of 10 psig as opposed to the current Technical Specification trip setting of 35 psig is only about 1 second. (Ref: FGLP-1905 dated 7/24/77). SAFETY LIMIT: None. With loss of reheat header pressure, the helium circulators can be driven by the auxiliary water turbine drives. There is no fuel failure or breach of the primary coolant pressure boundary. (Ref: FSAR Section 14.2.2). LIMITING INITIATING EVENT: Offset rupture of a hot reheat steam line or a cold reheat steam line in a section of line common to both loops. TRANSIENT OVERSHOOT: For the limiting initiating event, the reheat steam pressure drops to 10 psig within a few seconds. Further drop O

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   -   r            ,

Attachment 4 P-85214 ,c \ V ANALYSIS VALUE in pressure in the non-choke is significantly slower. (Ref: FGLP-1905 dated 7/24/77). TRANSIENT TIME RESPONSE: No delay assumed in sensing actual pressure. P_R0 CESS MEASUREMENT ACCURACY: Not applicable. DISCUSSION: FGLP-1905 dated 7/24/77 evaluated offset ruptures of reheat header piping in a number of locations including loop header piping as well as comon header piping. However, detection of steam leaks in reheat loop header piping within the reactor building is accomplished by the Steam Pipe Rupture Detection System consisting of ultrasonic noise detectors in conjunction with either the area pressure or temperature sensors. Reheat comon header offset ruptures for both cold and hot reheat 3 steam would be detected within a few seconds with reheat steam (V pressure at the sensors being reduced to near atmospheric pressure. Detection is not significantly affected by a reduction in the analysis value from 35 psig to 10 psig. The maximum change in detection time is only about 1 second. There is one exception where the hot reheater header pressure-low will not detect an offset rupture of reheat common header piping. This is for an offset rupture in hot reheat header piping downstream of the pressure sensors. In this case, hot reheat steam will continue to be supplied to the break at about 624 lb/sec and the pressure will be about 80 psia. A scram will probably result from circulator speed - feedwater flow two loop trouble. The derivation of the choke pressure for a reheat system rupture is contained in the attached calculation sheet by R. C. Potter. O

Attachment 4 P-85214 o L) ANAli' SIS VALUE CALCULATION SHEET CALCULATION FOR: Choke Pressure For Reheat System Rupture PROJECT: FSV PAGE 1 of 1 PREPARED BY: R. C. Potter DATE: 10/9/84 REF. DOCUMENTS: (1)ASMEPaperNo.63-WA-19 (2) GA Memo FSVA:246:77 OBJECTIVE: Determine the approximate choka pressure for a rupture in the F5V reheat system. According to reference 2 the PPS setpoint for. low hot reheat steam pressure should be set to a value above the choke pressure. The following curve (fromreference1)presentsthechokepressure ratio (exhaust over inlet pressure ratio) for steam. ._ l l bk k  ! _b f~ Jl Nlim I

          =.  .

l I II-----' I For 'the FSV Feheat sistem the pressures and temperaiu~r'es are: 25% Load 100% toad CRH pressure, psia 144 634 (reheat inlet) CRH temperature, degree F 441 670 (reheat inlet) HRH pressure, psia 129 568 (reheat outlet) HRH temperature, degree F 930 1000 (reheatoutlet) From the curve the choke pressure ratio can vary from ~.544 to ~.548 With an atmospheric exhaust pressure of 12.3 psia P choke = 12.3/.544 = 22.6 psia P choke = 12.3/.548 = 22,4 psia 22.6-12.3 = 10.3 psig 22.4-12.3 = 10.1 psig O '

Attachment 4 P-85214

 'O ANALYSIS VALUE FUNCTION: Reactor Scram FUNCTIONAL UNIT: Main Steam Pressure - Low ANALYSIS VALUE:    1500 psig PRESENT TECHNICAL SPECIFICATION TRIP SETTING: > 1500 psig LICENSING BASIS:      Low superheat steam pressure is an indication of main steam line rupture.      Immediate shutdown of the reactor is appropriate in such a situation. In addition, the superheater outlet stop check valves are automatically closed to reroute main steam to the flash tank (through the individual loop bypass valves and desuperheat'ers). This is necessary for the continued operation of the helium circulators on steam in the absence of the auxiliary boilers. The trip point is selected to be below normal operating levels and system transients (Ref:     FSAR Section 7.1.2.3).

SAFETY LIMIT: None in terms of main steam pressure which decays to atmospheric by subsequent actions initiated by the trip. Rerouted steam is still available to drive the helium circulators. There is no fuel failure nor breach of the primary coolant pressure boundary. LIMITING INITIATING EVENT: Offset rupture of a main steam header in ! the turbine building from 100% power operation. TRANSIENT OVERSHOOT: Pressure in the main steam rapidly decays to ( the analysis value of 1500 psig in 4 seconds. The pressure further falls to 1110 psig in the next 2 seconds after which the steam pressure and flow is primarily dictated by closure of the main steam output stop check valves HV-2223 and HV-2224. (Ref: Internal GA 1967 corresponence). TRANSIENT TIME RESPONSE: No delay assumed in sensing actual pressure. PROCESS MEASUREMENT ACCURACY: Not applicable. DISCUSSION: Main steam pressure - low plant protective system trip setting is for detection and isolation for main steam line ruptures in the turbine building. Loop main steam piping in the reactor O l t

l Attachment 4 P-85214 ANALYSIS VALUE 2 building is monitored for line breaks by the Steam Pipe Rupture Detection System and thus is not further addressed here.

• The GA internal analysis documented in 1967 assumed normal action on the part of safety and non-safety systems. Turbine trip occurs at about 2000 psig nain steam pressure and reactor scram at 1500 psig.

The reactor scram also initiates closure of the main steam outlet stop check valves which effectively isolates the leak within 10 seconds. No analysis.was made without the above control and safety actions.

        )

b O l O

Attachment 4 P-85214 O ANALYSIS VALUE FUNCTION: Reactor Scram FUNCTIONAL UNIT: Plant Electrical System Loss ANALYSIS VALUE: 264 volts (66V tap setting) PRESENT TECHNICAL SPECIFICATION TRIP SETTING: None LICENSING BASIS: Plant electrical system power loss requires a scram to take maximum advantage of whatever cooling flow can be maintained. The accident of concern is loss of outside power, complicated by a turbine generator trip and a failure of both emergency diesel generators to start. The control system initiates reactor power reduction upon. turbine generator trip. Voltage levels of <264 volts

      -on two out of three 480 volt ~ essential buses indicate loss of outside power and turbine generator trip. The buses are subsequently energized upon start ~f  o    an emergency generator with a delay of no startup of the deisel O greater   than 35before generators)     seconds   (to initiation scram    allow automatic (FSAR Sections 7.1.2.3 and 8.2.5.3.2). If neither standby diesel generator is available, the Alternate Cooling Method may be used (FSAR Section 8.2.8).

SAFETY LIMIT: None. The purpose of this reactor scram is to mitigate the consequences of an accident outside the design basis of the FSV reactor plant, i.e., loss of outside electric power and turbine trip, combined with failure of both ' emergency diesel generators to start. The automatic scram reduces the heat output of the reactor core as quickly as possible in this unlikely event. LIMITING INITIATING EVENT: ' Loss of outside power, complicated by a turbine generator trip and a failure of both emergency diesel generators to start (FSAR Section 7.1.2.3). TRANSIENT OVERSHOOT: Until standby diesel generators or an alternate power source to the 480 volt essential buses is supplied, voltage

       'will continue to degrade.

TRANSIENT TIME RESPONSE: Scram will be initiated within 35 seconds. Response time of electrical equipment is insignificant as compared to this delay. 4 O

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

5 j i j Attachment 4- ! P-85214 i O q i i i i 4

ANALYSIS VALUE 1

i i 1 ! PROCESS MEASUREMENT ACCURACY: Not applicable. i" t > i. t i i j .' i e i i [- . l6 r 4 jz 4 k i r t f r i i. A 4 4. s-e i i b' 4 k i

i 1

!G I '. ~

       ......u,.___.._..__.,                  .____.c...,..__,..,_____.                                                         ...,_ _ _.

Attachment 4 P-85214 10 V ANALYSIS VALUE FUNCTION: Reactor Scram FUNCTIONAL UNIT: High Reactor Building Temperature (Pipe Cavity) ANALYSIS VALUE: 175 Degree F PRESENT TECHNICAL SPECIFICATION TRIP SETTING: < 325 Degree F LICENSING BASIS: High reactor building temperature is indicative of a pipe rupture within the building. A scram is required to decrease ossible damage to equipment. The trip is set to detect large leaks p(Ref: FSARSection7.1.2.3). SAFETY LIMIT: None. Steam pipe ruptures do not result in damage to reactor plant equipment or structures nor result in core damage (Ref: FSAR Section 14.5.1). Offset rupture of a steam line in the O V LIMITING INITIATING EVENT: reactor building with the reactor in power production. TRANSIENT OVERSHOOT: Variable dependent upon type of steam line and reactor power level. For an offset rupture of a cold reheat steam line from 100% reactor power, the temperature in the pipe cavity would be 200 degree F in about 1/4 minute and would ' peak to about 270 degree F at 2 minutes. For the same cold reheat steam line offset rupture, but at 25% of rated reactor power, the temperatures would reach 175 degree F at 3 minutes and remain at that value. Offset ruptures of main steam lines and hot reheat steam lines because they contain higher energy steam would result in higher temperatures being attained more quickly than the cold reheat steam pipe rupture analyzed. (Ref: GP-0100-P dated 7/18/79). TRANSIENT TIME RESPONSE: Variable dependent upon type of steam line and reactor power level. See discussion under " Transient Overshoot" above. PROCESS MEASUREMENT ACCURACY: Not applicable. DISCUSSION: GP-0100-P dated 7/18/79 is an evaluation of the instrument setpoint for Temperature Sensors TSH-93448 thru TSH-93453. These particular temperature sensors are part of the Steam Pipe Rupture Detection System. This analysis is also appropriate for (O 1 1 Attachment 4 P-85214 , (D v 4 ANALYSIS VALUE evaluating the adequacy for the High Reactor Building Temperature sensors. Specifically, the latter temperature sensors are TE-93472, 3, and 4 and their locations are near TSH-93451, 2 and 3 in the pipe cavity. The Analysis value of 175 degree F matches the lowest predicted temperature for a major steam line break with the reactor plant in the electrical production mode. 'The predicted 175 degree F temperature is for an offset rupture of loop cold reheat steam line. With the plant at 100% of rated load, this same pipe rupture would result in a predicted temperature of 270 degree F. Loop main steam line and hot reheat steam line offset pipe ruptures would produce higher temperatures due to the higher energy steam. O V l L i l I l i

Attachment 4 P-85214 O ANALYSIS VALUE FUNCTION: Loop Shutdown FUNCTIONAL UNIT: Steam Pipe Rupture Under PCRV, Loop 1 Steam Pipe Rupture Under PCRV, Loop 2 Steam Pipe Rupture, North Pipe Cavity Loop 1 Steam Pipe Rupture, South Pipe Cavity Loop 1 Steam Pipe Rupture, North Pipe Cavity Loop 2 Steam Pipe Rupture, South' Pipe Cavity Loop 2 ANALYIS VALUE: 9 volts d.c. PRESENT TECHNICAL SPECIFICATION TRIP SETTING: < 9 volts d.c. LICENSING BASIS: The purpose of the ultrasonic detectors is to identify the specific secondary coolant loop within the reactor building containing a pipe rupture. The ultrasonic detectors analysis.value is established twice normal background level (FSAR Section 7.1.2.4 and Figure 7.1.-7). SAFETY LIMIT: None. If the ultrasonic detectors did not perform their safety function, reactor building temperature / pressure would increase and alarm in the control room. The operator may then use other instrument parameters to determine which loop should be isolated. Harsh environments in the reactor building, for purposes of environmental qualifications of equipment, assumed the steam leak was terminated manually by the operator 4 minutes after the onset of the event (GA-A-12045). I LIMITING INITIATING EVENT: A significant steam leak, defined as any steam leak that if unisolated would result in reactor building temperatures in excess of that established in Gulf-GA-A12045 report l Equi for Steam

     " Qualification of Fort St. Vrain Safe ShutdownEnvironment Resulting from Pipe Rup l

v Attachment 4 P-85214 O ANALYSIS VALUE Not discussed -in the FSAR.- Any significant

       -TRANSIENT- OVERSHOOT:

initiating steam leak would flood the detector with a signal much larger than twice the normal background noise level. TRANSIENT TIME RESPONSE: The response time to identify the leaking loop is 25 msec (F5AR Section 7.3.10). A modification is in process so that following loop identification there will be a 1 second delay to ensure ultrasonic sound source is not transitory. PROCESS MEASUREMENT ACCURACY: Not applicable. DISCUSSION: The pipe rupture detection system automatically detects a steam line rupture in the reactor building. It alarms and initiates loop shutdown using a combination of ultrasonic detectors, differential pressure and temperature switches. The ultrasonic detectors establish in which of the two loops the leak exists and the differential pressure or temperature switches confirm that the leak

      - is large enough for automatic loop shutdown action to take place.

The automatic action initiated is loop shutdown. This involves tripping of the two helium circulators and shutting off feedwater and steam to and from the steam generator of one loop. Loop shutdown is described in Section 7.1 of the FSAR. For pipe ' ruptures, it shuts off the source of feedwater/ steam to the leak. I i Ultrasonic detectors are located under the'PCRV, in the steam piping area at the south end of the reactor building and on the north wall of the reactor building. The trip point in each detector is set at 9 volts d.c. which is twice the normal background signal in the reactor building. This background setpoint was selected by testing to

        ; determine the highest spurious ultrasonic noise generated in the reactor building by such actions as pneumatic valves unloading.

A steam pipe. rupture rate greater than 20 pounds /sec. in either loop will be detected by the pressure sensors and ultrasonic detectors. A steam pipe rupture less than 20 pounds /sec. will be detected by the temperature sensors and the ultrasonic detectors. Also, the reactor

                                                    ~

building temperature switches can trip on high temperature and the reactor- building to atmospheric pressure differential pressure instruments can trip on high pressure. The ultrasonic detector signal from the leaking loop ~ will combine with either the high temperature or high differential pressure signal and trip the leaking Attachment 4 i P-85214 3 O ANALYSIS VALUE loop. Ultrasonic detectors monitoring the loop without a pipe rupture may eventually trip. However, first-in-with-lockout logic will prevent selection of the non-leaking loop and subsequent loop shutdown'and isolation of that loop. A -more detailed discussion of the pipe rupture detection system can be found in Section 7.3.10 of the FSAR. O l

  -O T
    ,-   e   .---s -.---,.wwc-.,.w.---4 .  . . , . ,   .,.,..,.,,,_._%,      , _ . _ _ _ , _ _ , _ , . , , , , , , . _ , , , , , ,   , . , _ , . _ _ _ . , , , , . _ , ,     . , _ _ , _ _ . , ,   _ _ _ , . , _ _ _ , , _ , , . .

Attachment 4 P-85214 ANALYSIS VALUE FUNCTION: Loop Shutdown FUNCTIONAL UNIT: 'High Pressure, Pipe Cavity High Pressure, Under PCRV ANALYSIS VALUE: 2.5" w.g. PRESENT TECHNICAL SPECIFICATION TRIP SETTING: < 2.5" w.g. LICENSING BASIS: The analysis value setpoint of 2.5" w.g. is above nomal reactor building pressure of - 1/4" w.g. but, below the pressure of about 3" w.g. at which the reactor building louvers open to relieve any overpressure condition whether resulting from a steam pipe rupture or a primary coolant depressurization accident (FSAR Section 6.2.3.1). SAFETY LIMIT: None. The reactor building overpressure protection Q system will relieve the overpressure transient should it occur. LIMITING INITIATING EVENT: Offset rupture of loop secondary steam piping in the reactor building. TRANSIENT OVERSHOOT: Variable dependent upon size of the steam leak. Maximum overshoot could be 10" w.g. above the refueling floor. The reactor building louver overpressure protection system is designed to limit overpressure transients above the refueling floor. TRANSIENT TIME RESPONSE: No delay assumed in measuring pressure. PROCESS MEASUREMENT ACCURACY: Not applicable. DISCUSSION: See Steam Pipe Rupture Ultrasonic Detector " Discussion." High pressure trip setting is part of the total steam pipe rupture detection system. O

Attachment 4 P-85214 bq . ANALYSIS VALUE FUNCTION: Loop Shutdown FUNCTIONAL UNIT: High Temperature, Pipe Cavity ANALYSIS VALUE: 130 degree F PRESENT TECHNICAL SPECIFICATION TRIP SETTING: < 130 degree F LICENSING BASIS: The analysis value is established to be above the normal ambient temperature in the pipe cavity 100 degree F is the ambient temperature per FSAR Table 7.1-3. The analysis value must be low enough to assure a fast response to steam pipe ruptures in the pipe cavity. Major steam line breaks from 100% power produce temperatures above 200 degree F in less than 1/4 minute (GA-A12045, Figure 3.11). A cold reheat steam pipe offset rupture from 25% reactor power produces the analysis value temperature of 130 degree F

     ~in less than 1/2 minute (Ref. GP-0100-P dated 7/18/79). These p response times are less than the 4 minutes assumed to define the V environmental qualification temperature envelope used to qualify safe shutdown cooling equipment in the reactor building. The design basis steam line break was assumed in this case to be an offset rupture of a cold reheat steam pipe from 100% of rated power.

SAFETY LIMIT: None. If the pipe cavity temperature sensors fail, the overpressure portion of the protection system will sense reactor building overpressure and provide inputs to the PPS along with the loop specific inputs provided by the ultrasonic detection system. LIMITING INITIATING EVENT: Offset rupture of loop secondary steam piping in the pipe cavity. TRANSIENT OVERSHOOT: Variable dependent upon size of the steam pipe and size of the break. Assuming an offset rupture of a cold reheat steam line and 100% power operation the temperature around the PCRV would peak to 200 degree F in less than 1/4 minute (GA-A12045, Figure 3.11). TRANSIENT TIME RESPONSE: No delay assumed in sensing actual temperature. O

>                                                                                                  Attachment 4 i                                                                                                 P-85214 ANALYSIS VALUE PROCESS MEASUREMENT ACCURACY: Not applicable.

DISCUSSION: See Steam Pipe Rupture Ultrasonic Detector " Discussion." High Temperature in the pipe cavity is part of the total steam pipe rupture detection system. I .i O 1: l { l t O i { l

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

                                                                ' Attachment 4 P-85214 O

L/ ANALYSIS VALUE FUNCTION: Loop Shutdown FUNCTIONAL UNIT: High Temperature, Under PCRV ANALYSIS VALUE: 130 degree F PRESENT TECHNICAL SPECIFICATION TRIP SETTING: < 130 degree F LICENSING BASIS: The analysis ~value is established to be above the normal temperature inside the PCRV support ring to preclude spurious trips. The ambient temperature under the PCRV is specified to be 100 degrees F in FSAR Table 7.1-3. Operating data indicate the ambient temperature can and at times does somewhat exceed the stated 100 degrees F. As illustrated in the discussion under " Transient Overshoot", the analysis value of-130 degrees F is exceeded within 10 seconds from the onset of a major steam pipe rupture, q SAFETY LIMIT: None. If the temperature sensors f' ail, the k / overpressure portion of the protection system will sense reactor building overpressure and provide inputs to the PPS along with the loop specific inputs provided by the ultrasonic detectors. LIMITING INITIATING EVENT: Offset rupture of a loop secondary steam piping under the PCRV. TRANSIENT OVERSHOOT: Variable dependent upon type of steam pipe, size of the break and location of the sensor from the steam pipe rupture. A limiting case regarding ability to detect would be a cold reheat steam pipe rupture (offset rupture assumed) occuring near one side within the PCRV support ring and the temperature sensor located at the far side (36.2 feet away). The resultant temperatures at the l I sensor located 36.2 feet away for both 100% and 25% power operation are as follows: O L

Attachment 4 P-85214 .O ANALYSIS VALUE' Tem)erature in degree F Time @ 100% Jower @ 25% Power 10 seconds 260 146 30 seconds 275 162 1 minutes 290 172 2 minutes 308 186 3 minutes 308 195 4 minutes 300 197 From the above it can be seen that offset steam pipe ruptures from 100% are detected almost instantly and from one quarter load within 1 minute. In both cases, this is much less than a cold reheat pipe offset rupture from 100% reactor power not isolated for 4 minutes which is the basis for environmental qualification of safe shutdown equipment in the reactor building. TIME RESPONSE: No delay assumed in sensing actual O TRANSIENT temperature. PROCESS MEASUREMENT ACCURACY: Not applicable. DISCUSSION: See Steam Pipe Rupture Ultrasonic Detector " Discussion." High temperature under the PCRV is part of the total steam pipe rupture detection system. O Attachment 4 P-85214 b3 ANALYSIS VALUE FUNCTION: Loop Trip FUNCTION UNIT: Steam Generator Penetration Over Pressure Loop 1 Steam Generator Penetration Overpressure Loop 2 ANALYSIS VALUE: 810 psig PRESENT TECHNICAL SPECIFICATION TRIP SETTING: < 810 psig LICENSING BASIS: The steam generator interspace rupture discs are set at 825 psig (nominal). The burst pressure range (+2%) is 809 psig to 842 psig (Tech Spec LSSS 3.3 Table 3.3.1). The relief valve is sized to allow a 370 psi pressure drop in a safety valve inlet line when the valve is relieving a nameplate capacity of 126,000 lb/hr superheated steam at 1000 degree F (50-11-6 Section 7.1). This prevents the penetration pressure from exceeding the reference O Pressure of 84s The enaissis vaiue of 81o Psis essures that Psisgenerator the affected steam loop will be isolated, which minimizes the quantity of steam to be relieved and adds assurance that the PCRV reference pressure will not be exceeded. It does not preclude that the relief train will be actuated. SAFETY LIMIT: 845 psig which is the Reference Pressure of the PCRV (F5AR Section 5.2.1.1 and Technical Specification Safety Limit 3.2). LIMITING INITIATING EVENT: Offset rupture of one steam or feedwater tube (subhender) in one penetration (FSAR Section 5.8.2.5.4). Failure of reheat steam piping within the penetration is not a penetration pressurizing source since reheat steam pressure is maintained below primary coolant and penetration interspace pressure. TRANSIENT OVERSHOOT: Variable dependent upon size of leak and pressure at which the rupture discs rupture. Could be up to 842 4 psig. TRANSIENT TIME RESPONSE: No delay assumed in sensing actual penetration pressure. I

Attachment 4 P-85214 O. ANALYSIS VALUE PROCESS MEASUREMENT ACCURACY: Not applicable. DISCUSSION: Steam Generator penetration overpressure is indicative of a main steam subheader or feedwater subheader rupture within a penetration. All other high energy lines within penetrations contain steam / water below the pressure at which the penetrations are

  . maintained. Steam   generator penetration pressure is normally maintained at about 10 psi above primary coolant pressure.      However, due to some leakage in reheat steam piping within the penetrations, some penetrations can be and are maintained at pressures slightly exceeding reheat steam pressure to limit the quantity of purified helium leaking into the reheat steam (FSAR Section 5.8.2.5.4).

The steam generator penetration overpressure PPS trip results in a loop trip which isolates high energy sttam/ water from the leak. The redundant relief trains assure that the steam generator penetrations do not exceed the PCRV reference pressure of 845 psig. O O Attachment 4 P-85214 O v ANALYSIS VALUE FUNCTION: Loop Shutdown FUNCTIONAL UNIT: High Reheat Header Activity, Loop 1 and Loop 2 ANALYSIS VALUE: 5 mrem /hr Above Background PRESENT TECHNICAL SPECIFICATION TRIP SETTING: -

                                                           <5 mrem /hr Above

Background

LICENSING BASIS: High reheat header activity is an indication of a reheater tube rupture resulting in leakage of reactor helium into the reheat steam system (FSAR Section 7.1.2.4). The analysis value assures detection of major reheat tube ruptures and an on scale reading with up to design value circulating activity for post accident monitoring. Detection of smaller size leaks or with low circulating coolant activity can be detected and alarmed by the backup reheat condensate monitors and/or the air ejector monitor q (Reference FSAR Section 7.3.5.1). C' SAFETY LIMIT: None. The primary coolant pressure boundary has already been breached in the form of a failure of a reheater tube. The reheat header activity monitors provide detection and automatic loop isolation for significant size ruptures. LIMITING INITIATING EVENT: Offset rupture of a reheater tube with subsequent inleakage of primary coolant into the reheat steam system. TRANSIENT OVERSHOOT: Variable dependent upon size of leak, circulating activity and reactor power level with low power increasing the leak rate due to lower reheat steam pressure. The 5 mr/hr analysis value is equivalent to 1 Ci/sec leak rate which leaves 5 Ci/sec for an on-scale recording of an overshoot (FSAR Section 7.3.5.1). TRANSIENT TIME RESPONSE: A step increase in leakage from normal to a will be detected in value corresponding less than (to above the trip setting a second FSARSection7.3.5.1). PROCESS MEASUREMENT ACCURACY: Not applicable DISCUSSION: The hot reheat steam lines are monitored for possible activity leakage into the steam system resulting from a leaking O .

Attachment 4 P-85214 O ANALYSIS VALUE e reheater tube. The activity monitors provide for continuous monitoring from about 0.1 Ci/sec to the maximum possible release rate of about 6 Ci/sec. The activity monitors are set to trip at 5 mr/hr above background which corresponds to a release rate of about 1 C1/sec. These detectors are located adjacent to the loop reheat header and are shielded against background radiation by 1-1/2 inch of lead. They are apart of the plant protective system with the associated electronics located in control room control board I-10. A detailed discussion of the reheat steam activity detection system is contained in FSAR Section 7.3.5.1. The total detection system includes the loop header radiation monitors, the loop header condensate monitors and the air ejector monitor. Only the loop header radiation monitors can initiate an automatic loop shutdown for isolation of the leak. The other monitors provide for detection of smaller size leaks for which operator manual action is appropriate. O l O

Attachment 4 P-85214 O 1 ANALYSIS VALUE FUNCTION: Loop Shutdown FUNCTIONAL UNIT: Low Superheat Header Temperature, Loop 1 Low Superheat Header Temperature, Loop 2 High Differential Temperature Between Loop 1 and Loop 2 ANALYSIS VALUE: Low temperature Loop 1 and 2, 780 degree F High Differential Between Loops, 65 degree F PRESENT TECHNICAL SPECIFICATION TRIP SETTING: Low temperature Loop 1 and High2,diTferential> 800 degree F between loops < 50 degree F LICENSING BASIS: The Low Superheat Header Temperature trip is to g protect the steam generator from a floodout transient. Low superheat

 \ header temperature in a loop is indicative either of a feedwater valve or controller failure yielding an excessive loop feedwater flow rate or a deficiency of helium flow, and a loop shutdown is in order for steam generator protection (FSAR Section 7.1.2.4). The required coincident high differential temperature between loops functions to prevent the loop trip from occuring during normal operation at low main steam temperatures such as in a normal plant shutdown. The analysis values were selected to be sufficiently above                 steam generator floodout temperature which is 660 degree F at 2400 psig and sufficiently below normal operating temperature which is at a nonnal minimum of 880 degree F at 30% power. The differential temperature between loops analysis value of 65 degree F assures a loop trip before a turbine / generator trip. See Discussion" for additional details. (Ref: GP-0368 dated 5/13/80).

SAFETY LIMIT: None. Low superheat header temperature does not result in fuel failure or a breach of the primary coolant pressure boundary. LIMITING INITIATING EVENT: Either excessive feedwater flow rate or deficient helium flow rate. TRANSIENT OVERSHOOT: Not specifically analyzed. Once the loop trip occurs, the steam generator is isolated and the helium circulators O

  .j?

Attachment 4 P-85214 {- ANALYSIS VALUE are shutdown. At that time the steam generator and superheat temperatures would stabilize with a continuing gradual decrease in temperature. TRANSIENT TIME RESPONSE: The thermocouples and thermocouple wells will cause a time lag. This time lag was not considered in the low superheat header temperature evaluation because the time constant for the steam generator to respond to feedwater/ helium flow rate changes is about 30 _ seconds. The low circulator speed programmed by feedwater flow rate is more ideally suited to handle fast transients (GP-0095 dated 7/5/79 and P-80340 dated 10/1/80). PROCESS MEASUREMENT ACCURACY: Not applicable. DISCUSSION:- The low superheat header temperature loop trip is designed to protect against malfunctions causing either excessive feedwater flow rate in a loop or deficient helium flow rate in a n loop. Coincident low superheat steam temperature and a temperature (/ mismatch between the loops is required to produce the loop trip. The trip is only active when the Interlock Sequence Switch is in the power position. The trip is disabled on a reactor scram since main steam temperature will drop as designed. A non-uniform rate of temperature decrease between the two steam generator loops after a ! reactor scram does not indicate a malfunction and therefore is not a 1 concern. i An additional function of the low superheat header temperature loop trip is that it protects against turbine / generator trip on low main steam temperature. The turbine / generator low main steam temperature i trip setpoint is 800 degree-F. The main steam from the two steam generator loops is mixed upstream of the turbine inlet temperature sensors. The following would be steam temperatures at the time that analysis value loop trip conditions would be attained for 30 and 100% of rated power operation. ! 30% Power l [ Malfunctioning Loop Temp: 780 degree F Other Loop Temp: 880 degree F Turbine Mixed Inlet Steam Temp: 830 degree F Loop-to-Loop delta T: 100 degree F IO i L

Attachment 4 P-85214 0 . ANALYSIS VALUE 100% Power Malfunctioning Loop Temp: 780 degree F Other Loop Temp: 1000 degree F Turbine Mixed Inlet Steam Temp: 890 degree F Loop-to-Loop deltnT: 220 degree F From the above it can be seen that conditions for trip of a malfunctioning loop are attained prior to reaching the turbine trip setpoint. Therefore, turbine trip is precluded due to a single malfunctioning loo over the plant power operating range. (Ref: GP-0368 dated 5/13/80 . O L i 1 i i l 0 . t I

Attachment 4 P-85214 O ANALYSIS VALUE FUNCTION: Circulator Trip FUNCTIONAL UNIT: Circulator Speed-Low ANALYSIS VALUE: 2390 rpm below normal as programed by feedwater flow. (Ref: GP-0264 dated 2/15/84 and P-80340 dated 10/1/80). PRESENT TECHNICAL SPECIFICATION TRIP SETTING: 1910 rpm below normal as programmed by feedwater flow. LICENSING BASIS: Too low a circulator s causes a mismatch between thermal power input and heat removal (peed feedwater flow) in a steam generator, which may result in flooding the superheater section. The circulator trip causes an automatic adjustment, as required, .in the turbine governor setting, feedwater flow, and remaining' circulator speed to maintain stable steam pressure and temperature conditions. FSARSection7.1.2.6). O SAFETv uMIT: None. tow circuiator speed, if not corrected, couid result in floodout of a steam generator loop. While this is an undesirable temperature transient, it will not result in a breach of the primary coolant pressure boundary. In addition, the programed low circulator speed trip is backed up by low main steam temperature trip in conjunction with steam temperature mismatch between steam generator loops which produces a loop trip. LIMITING INITIATING EVENT: A sudden loss of circulator motive power l with the circulator coasting to self-turbining speed. ! TRANSIENT OVERSHOOT: Varied dependent upon the malfunction producing l the low circulator speed. On total loss of motive power the circulator. will coast down to self-turbining speed of 700 rpm in

about 30 seconds. (FSARSection4.3.1and4.2.2.3.2)

TRANSIENT TIME RESPONSE

A delay of 5 seconds is incorporated in the

instrument channel to discriminate against short term transients ~ in

, circulator speed. (FSAR Section 4.3.1; GP-0112

P-80340)

PROCESS MEASUREMENT ACCURACY: Not applicable. l ' !- DISCUSSION: Low circulator speed is an indication of a speed control l or other equipment failure and results in a decrease in loop helium O i l L

Attachment 4 P-85214 Il v ANALYSIS VALUE flow. This leads to a mismatch between heat input and heat removal (feedwater flow rate) in a steam generator. Such a mismatch can result in a low superheat header temperature and a high differential temperature between loops. This condition causes a PPS loop trip to prevent steam generator main bundle floodout.

      .To minimize loop trips when one circulator is malfunctioning, the PPS trip on low circulator speed has been incorporated. The circulator trip initiates a power runback to 50% (if the initial power is greater than 50%), and the remaining circulator in the loop is given a step change in program speed to maintain the required loop helium flow. Thus, two-loop plan operation can be maintained at up to 50%

power. Circulator coast down characteristics are such that circulator malfunctions are detected quickly on the basis of a speed measurement. U p coast down 25% (ponfrom 2390 rpm) complete rated speedloss of driving is 2 seconds; at partpower, load PPS action has an intentional the time taken v the time would be up to 4 seconds. delay of 5 seconds to discriminate against transient speed deviations. The time constant of the steam generator superheater header temperature responding to a change in helium flow is 30 seconds. Therefore, a reduction in circulator speed approximately(analysis of 2390 rpm value) in less than 30 seconds willresult in the trip of a single circulator followed by a power runback (if required) and a speedup of the remaining circulator, thus avoiding a loop trip on low superheat temperature. i (o l.

Attachment 4 P-85214 O ANALYSIS VALUE FUNCTION: Circulator Trip

 ' FUNCTIONAL UNIT: Loop 1, Fixed Feedwater Flow-Low (Both Circulators)

Loop 2, Fixed Feedwater Flow-Low (Both Circulators) ANALYSIS VALUE: 0% of rated full load (FGLP.P-1900 and'P-80340). 5% of rated full load is recommended as an~ analysis value to ensure an on-scale set point. PRESENT TECHNICAL SPECIFICATION TRIP SETTING: 20% of rated full load. LICENSING BASIS: The trip setting is below the normal operation range of 25% to 100% feedwater flow rate for which this trip is A activated- with the ISS switch in the power position. The trip setting is also high enough to preclude unstable steam generator flow conditions which were anticipated below about 20% feedwater flow rate. SAFETY LIMIT: None. Complete loss of feedwater flow prior to circulator trip has been shown to be acceptable. Unstable flow conditions in the steam generator would not cause damage even if persisting for a long period of time. LIMITING INITIATING EVENT: Sudden feedwater isolation at the feedwater end of.a steam generator. TRANSIENT OVERSHOOT: With either sudden steam generator isolation or a pipe rupture, the steam ~ generator flow would drop to zero. Therefore, using the recommended 5% of rated flow rate as the analysis value, the overshoot would be 5% of rated flow rate. TRANSIENT TIME RESPONSE: The PPS circuitry includes a 5 second delay during which the feedwater flow rate must be maintained below the setpoint for the trip to take. place. This protects against short ter:n system transients for which a trip is not needed or desirable. The 5 second delay is shown'on FSAR Figure 7.1-16 and IB-93-8. O

Attachment 4 P-85214 O ANALYSIS VALUE PROCESS MEASUREMENT ACCURACY: Not applicable. DISCUSSION: The analysis presented in FGLP-1900 and attachment 6 to P-80340 dated 10/1/80 assumed a feedwater valve closure with the feedwater flow rate then reduced to zero in 3 seccnds. The time for the steam generator tubes to reach ASME code allowable temperatures including consideration for the hottest tube in the bundle was determined. Credit was taken for reduced pressure in the steam generator tubes since this permits higher code allowable temperatures. For this same reason a feedwater line break is not a limiting condition since the line break permits a dump of the steam i generator and lower pressures within the steam generator tube. The most limiting case is a feedwater valve closure which stops feedwater flow to one steam generator loop from two loop full power operation. In this case code allowable temperatures are reached at 6.6 seconds and occurs at Superheater I outlet. This is in excess of the 5 second delay which is part of this PPS circuitry and therefore is All other conditions investigated which include two loop C- acceptable. 25% power operation and single loop 50% power operation had much larger margins. While it has been concluded that unstable steam generator flow conditions which can occur below 20% of rated feedwater flow rate do' r.ot constitute a safety limit, it is desirable to avoid unstable flow r conditions for prolonged periods of time. It was previously recommended, and is still recommended, that an alarm at 20% of rated flow rate be incorporated to alert the operator to this condition. l O Attachment 4 P-85214 O ANALYSIS VALUE FUNCTION: Circulator Trip FUNCTIONAL UNIT: Loss of Circulator Bearing Water ANALYSIS VALUE: 450 psid. (GP-C211, dated December 12, 1979 and P-80340,datedOctober1,1980). PRESENT TECHNICAL SPECIFICATION TRIP SETTING: > 475 psid LICENSING BASIS: In order to prevent circulator damage upon loss of normal and backup bearing water supplies, a gas pressurized water accumulator is fired when bearing water pressure falls below a trip setpoint value. The setpoint value is selected so that adequate water pressure is available during circulator coast down, which lasts for about 30 seconds, to maintain clearances within the circulator bearings of at least 0.001 in. Tests and analyses have shown that a trip at 450 psid provides substantial clearance margin about 0.001 in, when the circulators are operating at normal speeds. SAFETY LIMIT: None. Tripofacirculatorinitiatesaction(e.g., reduction reactor power) to maintain acceptable plant operating conditions. The affected circulator is safely brought to 700 rpm at which time the brake is set and the seal is set approximately 30 seconds later. LIMITING INITIATING EVENT: Loss of normal bearing water with failure to initiate backup bearing water supply (i.e., feedwater supply). TRANSIENT OVERSHOOT: 25 to 50 psid shortly after setpoint is reached before the fired accumulator restores the bearing water differential pressure (GP-0211andP-80340). TRANSIENT TIME RESPONSE: No delay assumed in sensing the bearing water differential pressure. PROCESS MEASUREMENT ACCURACY: Not applicable. DISCUSSION: Normal operating clearances of the journal and thrust bearings range from 0.0025 in. to 0.0045 in. A clearance of 0.001 in. has been selected as a conservative minimum clearance to assure adequate lubrication during a circulator shutdown from 100% speed. Measured minimum bearing pressures when the accumulator was fired at O Attachment 4 P-85214 m V ANALYSIS VALUE 475 psid ranged from 405 to 375 psid for the four circulators. If the accumulators had fired at 450 psid, it can be deduced that the minimum bearing pressure would be 350 psid. Loads on the journal bearing are due to imbalance of the rotor. Prior to circulator assembly, the rotor is balanced so that residual static imbalance and dynamically imposed imbalance due to coupling moments is less than 0.2 inch-ounces in any plane. This gives a displacement of 0.00004 in, at 10,800 rpm and a water delta P of 700 psi. Under the same conditions, but with a delta P of 350 psi, the shaft displacement will be only 0.00008 in. Tests have shown that the circulators can be shut down from 8000 rpm with a bearing pressure of only 50 psid. The thrust bearings also have substantial design margin with respect to bearing water pressure required to maintain a clearance of 0.001 in. The reverse thrust bearing carries a load of 2000 lbs. at design A operating speed. The maximum load, 2900 lbs. , would occur during a V rapid PCRV depressurization while the circulator is at 100% speed. For this reverse thrust load, a bearing water pressure of 263 psid will maintain a clearance of 0.001 in. The main thrust bearing carries a maximum load of about 8000 lbs. at 30% of rated speed. This bearing is designed for a maximum thrust of 11,400 lbs. At a load of 11,400 lbs. and a circulator speed of 2000 rpm, 350 psid bearing water pressure will maintain a clearance of 0.001 in. The clearance improves with increasing circulator speed. (Ref: GP-0211 and P-80340) O V Attachment 4 P-85214 m

  ,U ANALYSIS VALUE
                ,                          t FUNCTION: Circulator Trip FUNCTIONAL UNIT:       Circulator Penetration Trouble ANALYSIS VALUE: 810 psig (FSAR Table 7.1- 4)

PRESENT TECHNICAL SPECIFICATION TRIP SETTING: ff Bi0 psid 1 LICENSING BASIS: Penetration interspace rupture discs are set at 825 psig tyominal). The burst pressure range (+ 2%) is 809 psig to 842 psig.l(TechSpecLSSS3.3 Table 3.3.1). The relief valve is sized to allow a 40 psi pressure drop in the safety Vlave inlet line when the valve is relieving at nameplate capacity (170 gpm). Ref: S0-11-6 Section 7.1. Thus, an analysis value for circulator penetration trouble of 810 psig for circulator trip and isolation assures that the PCRV will not exceed the reference pressure of 845 psig. The analysis value of 810 piig does not preclude that the relief train will be activated. SAFETY LIMIT: 845 psig which is the Reference Pressure of the PCRV. (F5AR 5ection 5.2.1.1 and Tech Spec Safety Limit 3.2) LIMITING INITIATING EVENT: Bearing water leak of about 170 gpm within penetration. (F5AR Sections 4.2.2.3.2 and 4.2.2.3.7). TRANSIENT OVERSHOOT: Variable dependent upon size and location of bearing water line break and operator actions assumed (FSAR Sections 4.2.2.3.7 and 5.8.2.5.5). In all cases the redundant pressure relief trains assure that PCRV reference pressure of 845 psig will not be exceeded. TRANSIENT TIME RESPONSE: No delay assumed in sensing actual penetration pressure. FR0 CESS MEASUREMENT ACCURACY: Not applicable. DISCUSSION: Circulator penetration overpressure is indicative of a

,     bearing water line rupture within the penetration. All other high energy lines contain fluids of less pressure than the penetration pressure which is maintained at about 10 psi above primary coolant pressure. 'Upon reaching the penetration high pressure trip setting the circulator is tripped which isolates bearing water after          a 30
                                                                                                                                  'I 1

Attachment 4 P-85214 0 . ANALYSIS VALUE second coast down of the circulators to self-turbining speed. Other system actions are possible and probable as discussed in FSAR Section 4.2.2.3.7. If the line break is in the normal bearing water line and is of sufficient size to cause the bearing water pressure to drop below 600 psid then there will be an automatic switch to the backup bearing water supply line. This is alarmed in the control room. However, the normal bearing water system will continue to supply bearing water to the break until isolated by the operator or when low water level in the bearing water surge tanks automatically isolate the normal bearing water supply. In all situations the redundant penetration relief trains are capable of relieving the over pressure, maintaining the penetration within the PCRV Reference Pressure of 845 psig. i 1 I r 1 i O i

         .-   ,,J..
                             - - , , . ,- -    ~,      . . . , . . , . . . . _ . . . . - - , . - . . , _ . . . , - . _ . . . . - . - . - _ . . _ , . - - - . , , , _ . , , - , . . _ -     --_.-.n.,

Attachment 4 P-85214 p ANALYSIS VALUE FUNCTION: Circulator Trip FUNCTIONAL UNIT: Circulator Drain Malfunction ANALYSIS VALUE: 5 psid (FSAR Section 4.2.2.3.5 and Table 7.1 - 4) PRESENT TECHNICAL SPECIFICATION TRIP SETTING: > 5 psid LICENSING BASIS: This trip is provided to prevent steam from entering the bearing of an operating circulator. A differential pressure controller is utilized to maintain the bearing water main drain pressure at least 45 psi above the steam turbine exhaust pressure. When the pressure differential drops to 35 psi, the steam water drain control valves are opened to prevent steam from entering the bearings. If the~ above controls do not work, three PPS differential pressure switches for each circulator, set at > 5 psid, will initiate an automatic shutdown of the circulator. P d (Ref. FSAR Sections 4.2.2.3.5 and 7.1.2.6) SAFETY LIMIT: None in terms of fuel failure or breach of the primary coolant pressure boundary. The trip prevents or limits steam from

     )assing across the lower seal to the main drain which could be larmful to circulator seal and bearings (50-21-2, Section 8.1.6).

LIMITING INITIATING EVENT: System malfunction resulting in steam / water drain pressure not being maintained below main drain pressure. TRANSIENT OVERSHOOT: Not identified. TRANSIENT TIME RESPONSE: No delay is assumed in sensing pressure. PROCESS MEASUREMENT ACCURACY: Not applicable. DISCUSSION: None required. , ,O. - v L

                                                                                      ~ Attachment 4 P-85214

O  :

ANALYSIS VALUE FUNCTION: . Circulator Trip FUNCTIONAL UNIT: Circulator Speed High-Steam ANALYSIS VALUE: 11,700 rpm (P-84137 dated 5/10/84). PRESENT TECHNICAL SPECIFICATION TRIP SETTING: 1 11,000 rpm LICENSING BASIS: The speed sensing system response and trip setting are chosen so that under the maximum overspeed situation possible

!       (loss of restraining torque) the circulator will remain within design capabilities (FSAR Section 7.1.2.6).

SAFETY LIMIT: None in ~ terms of fuel failure or breach of primary . coolant pressure boundary. The design overspeed of the circulator is 13,500 rpm (FSAR Table 4.2-2). A production circulator was tested up to 13,070 rpm at 180 psig O helium pressure and subsequent teardown indicated no change to the hardware. Each production brake and disc assembly was spin tested (16,000 rpm at room temperature) which is an overspeed representative of 13,500 rpm at reactor operating conditions. Further, from tests and analyses, the disc catcher has been shown to provide complete - containment of discs and blades for a failure at 190% (18,145 rpm) of rated speed (9550 rpm). (FSAR Section A.14.3.3. and A.14.5). LIMITING INITIATING EVENT: Loss of restraining torque (no compressor blades) and reheat steam pipe-ruptures. TRANSIENT OVERSHOOT: See " Discussion" for Details. TRANSIENT TIME RESPONSE: 50 milliseconds for solid state electronics (utilized in 1969 accident analyses performed by_GA). PROCESS MEASUREMENT ACCURACY: Not applicable. DISCUSSION: The two limiting accidents leading to circulator overspeed are loss of-restraining torque (blade shedding)'and reheat , steam line ruptures. For loss of restraining torque, two trip i settings were analyzed with both assuming a delay of 50 milliseconds due to solid state electrical circuitry with the following results: O f L

Attachment 4 P-85214 O ANALYSIS VALUE Overspeed Trip Setting Peak Speed Beyond Trip RPM RPM Setting PRM 11,000 13,050 2,050 11,500 13,267 1,767 While an analysis value trip setting of 11,700 rpm (P-84137) was not investigated by GA at the time the above analyses were performed in 1969, it can be reasoned that a trip setting of 11,700 would also give acceptable results. Note that a change in tirp setting of 500 rpm from 11,000 to 11,500 rpm results in only a 217 rpm in increased overspeed. This is due to limited driving power for the circulator. An incremental increase of 200 rpm in the analysis value trip setting

< to 11,700 rpm then at most should result in about one half that increment in circulator overspeed. The resulting peak overspeed then would be 13,360 - 13,370 rpm which is less than the design overspeed of 13,500 rpm and thus is acceptable.

Steam line ruptures downstream of the circulator turbines were investigated. For these analyses the overspeed trip setting was 11,000 rpm and a 50 millisecond delay was utilized with the following results from which the results of a trip at 11,700 rpm are deduced. Time Peak Speed Reached Peak Speed (Seconds) Failed Other Failed Other Loop Loop Loop Loop Cold Reheat Line Rupture: No Control Action or Trip 13,264 12,587 4.6 8.0 Normal Control 10,399 9,922 0.4 1.1 Trip Active but No Control 11,517 11,345 0.6 1.4 Attachment 4 P-85214 G \,J ANALYSIS VALUE Two Loops Hot Reheat Line Rupture: No Control Action or Trip 13,036 4.5 Normal Control 10,371 0.4 Trip Active but No Control 11,450 0.6 From the above it can be seen that with no control action or rip the design overspeed of 13,500 rpm is not reached. Further, with normal control, the trip setting is not reached. With the trip active but no normal control, the transient overshoots are not large. The above data support an analysis value trip setting of 11,700. rpm since the design overspeed of 13,500 rpm is not reached for reheat steam pipe ruptures assuming no control or PPS trip. O O Attachment 4 P-85214 p V , ANALYSIS VALUE FUNCTION: Circulator Trip FUNCTIONAL UNIT: Circulator Seal Malfunction i ANALYSIS VALUE: -10" H20, or +80" H20 PRESENT TECHNICAL SPECIFICATION TRIP SETTING: > 10" H20, or < 80" H20 LICENSING BASIS: The differential pressure across the buffer helium labryrinth seal is normally maintained at +30" H20. A high reverse differential of -10" H2O would be reasonable evidence that bearing water is leaking into the primary coolant system. An increasing differential pressure of +80" H20 would be reasonable evidence that primary coolant is leaking into the bearing water and thus into the closed circulator service system. In both cases a circulator trip with brake and seals set is appropriate. (FSAR Sections 4.2.2.3.5 and 7.1.2.6; Table 7.1-4). SAFETY LIMIT: None. Trips are to prevent leakage of bearing water into the primary coolant system and leakage of primary coolant helium into the closed circulator service system. No fuel failure or breach of the primary coolant system will occur. LIMITING INITIATING EVENT: No limiting event has been identified. Various failures or maloperations of valves, sensors, and controllers can result in trip setpoints being reached. TRANSIENT OVERSHOOT: Not identified. It would be variable dependent on the cause of the maloperation. TRANSIENT TIME RESPONSE: A 3 second delay is built-in to preclude trips on brief disturbances of the buffer. seal, (Ref: SD-21-2, pg. 27, and IB-93-7). No other delay in sensing differential pressure was assumed. PROCESS MEASUREMENT ACCURACY: Not applicable. DISCUSSION: The circulator seal malfunction circuitry contains a number of first-in-with-lockout features. A simultaneous malfunction indication of both circulators in a loop would be indicative of a control transient rather than a coincident seal failure. The trip of O Attachment 4 P-85214 (3 v ANALYSIS VALUE the circulator is limited to the first circulator to reach the trip setpoint. Also, if one circulator is tripped with the brake and seal set, the second circulator in that loop is inhibited from tripping on a seal malfunction indication to prevent a LOFC. The remaining operating circulator can be manually tripped at any time. Additionally, a circulator seal malfunction trip, which could have injected bearing water into the primary coolant system, inhibits on first-in-with-lockout a subsequent trip of the PPS low level DPMM. There is no inhibit for the PPS high level DPMM. Thus, if the resultant primary coolant moisture content increase > 500 ppmv (67 degreeF),theisolationanddumpof a secondary cooTant loop is avoided, but the reactor would scram on trip of both high level monitors. These actions are appropriate in that the source of water has been identified as coming from the affected circulator, rather than the steam generator, and the scram will rapidly reduce core temperatures and stop water / graphite reaction. l l l l 1 O  ! l

Attachment 4 P-85214 O ANALYSIS VALUE FUNCTION: Circulator Trip FUNCTIONAL UNIT: Circulator Speed - High, Water ANALYSIS VALUE: 8800 rpm PRESENT TECHNICAL SPECIFICATION TRIP SETTING: < 8800 rpm LICENSING BASIS: Original setpoint was 11,500 rpm but operation at 10,500 rpm with cast Inconel 718 water turbines resulted in fatigue cracks in the bucket areas. The water turbines were changed to forged Inconel 718 for greater resistance to fatigue and the setpoint lowered to 8800 rpm. This permits continuous operation on water drive at 8000 rpm, if required. SAFETY LIMIT: None in terms of fuel failure or breach of the primary coolant pressure boundary. Actually, the water turbine is not within 7 the primary coolant boundary (GA-A13175 dated 10/16/84, Section (d 2.3.5). While a specific safety limit for. water turbine transient over-speed when on water drive has not been established, a prudent limit can be inferred from operational experience at FSV. See

    " Discussion" for details of cracks discoved in water turbines and the corrective action. A prudent upper limit for transient over-speeds of a water turbine water drive is at least 10,200 rpm. The basis for this is that one cast water turbine was run at FSV at 10,200 rpm.

While this particular water turbine did have cracks in the turbine buckets, they were self-arresting. Additionally, three cast water turbines have been operated ~ up to 8500 rpm without exhibiting any bucket fatigue cracking. All the water turbines have been changed to forged Inconel 718 which has superior fatigue cracking resistance. The above information has been extracted from GA-A13175, including Supplement 1 " Investigation' Into Cause and Consequence of the FSV Pelton Wheel Incipient Fractures" dated 10/16/84 and 11/14/75. LIMITING INITIATING EVENT: High speed operation on water turbine drive is only possible with a depressurized PCRV and feedwater for the motive power to the circulator water turbine. This combination-is only' required for the Rapid Depressurization/ Blowdown Accident (DBA #2) and would generally be used only for that combination. Attachment 4 P-85214 ANALYSIS VALUE TRANSIENT OVERSHOOT: Not analyzed. TRANSIENT TIME RESPONSE: Not analyzed although the 50 milliseconds

    . delay for solid state electronic circuitry would apply.

PROCESS MEASUREMENT ACCURACY: Not Applicable. DISCUSSION: In mid-1974, cracks were discovered in a water turbine at F5V. This led to reevaluation of safety analyses utilizing lower circulator speeds when the circulators are operated on water drive. Only DBA -2 was impacted because only with a depressurized PCRV are higher speeds possible with the circulators driven by the water turbine. 7000 rpm, down from 10,500 rpm, was selected as the circulator speed for DBA -2 and acceptable accident consequences were obtained. Additionally, the original cast Inconel 718 water turbines were replaced with forged Inconel 718 water turbines for greater resistance to fatigue cracking. Endurance testing with forged water turbines up to 7000 rpm was performed and no cracking was observed. O ultimately, in License Amendment No. 9, the Commission approved the DBA -2 accident reanalysis and the use of forged Inconel 718 for the water turbines. This same Amendment also set the overspeed trip setting at < 8800 PPM so that the water turbines could be operated at 8000 rpm, iT required. The 8000 rpm speed on water turbine was subsquently used as the steady state operating speed in a further reanalysis of DBA -2. See P-77221 dated. 11/01/77 and FSV License Amendment No. 22 dated 08/19/80. This reanalysis was not related to the original water turbine cracking problem. l l r i l O V L:

Attachment 4 P-85214 ANALYSIS VALUE FUNCTION: Circulator Trip FUNCTIONAL UNIT: Feedwater Flow-Low ANALYSIS VALUE: 20% below normal programed with circulator speed (Ref. F5AR Table 7.1-4 and FSAR Figure 7.1-11)

   .PRESENT     TECHNICAL     SPECIFICATION    TRIP SETTING:    Not currently contained in the Tech Spec LICENSING BASIS:        Low feedwater flow indicated control or equipment failures and can result in an increase in superheater outlet temperature. A circulator speed controller failure can lead to a higher than normal circulator speeds and a corresponding increase in superheater steam outict temperature. The programed low-feedwater-flow circulator trip prevents prolonged operation in the region of speed    versus    flow which may cause excessive superheat steam g temperatures.         Separate   circulator speed versus feedwater flow U programs are required dependent upon whether one or two circulators are operating in a loop. 

Reference:

FSAR Section 7.1.2.6. SAFETY LIMIT: Not specifically quantified in the FSAR or by calculations performed by GA Technologies. Initial studies by GA indicate that with the plant control system in manual that a deviation of a single circulator speed high by 20.6% at 100% reactor power will cause long term steam generator tube ASME Code allowable temperature limits to be reached. This does not in itself mean the steam. gneerator tube (s) would fail. At lower reactor power levels, greater deviations in high circulator speed could occur prior to reaching ASME Code allowable tube temperatures limits. Similar studies with the plant control system in automatic have not been performed. Logic, however, would suggest that the plant control system would attempt to slow down the speed of any circulator still responding to control commands thus compensating for the malfunctions circulator. As such, larger deviations in speed of the malfunctioning circulator should be possible without exceeding code allowable steam generator tube temperatures limits. LIMITING INITIATING EVENT: Circulator control or equipment failure causing a circulator to operate at excessive ~ speeds with potential mismatch in primary coolant helium flow and secondary coolant feedwater flow. Attachment 4 P-85214 V ANALYSIS VALUE TRANSIENT OVERSHOOT: Not identified. TRANSIENT TIME RESPONSE: A delay of 5 seconds is incorporated in the instrument channel to discriminate against short term transients in circulator speed (Ref: FSAR Figure 7.1-16). PROCESS MEASUREMENT ACCURACY: Not Applicable. DISCUSSION: This particular Plant Protective System (PPS) Functional Unit would better be named " Circulator Speed-High, Programed" rather than "Feedwater Flow-Low". The protective action is a single circulator trip. Feedwater flow is a loop parameter and comon to both circulators in a loop. The margin between the normal operating value and the setpoint used to date has proved acceptable for plant operations. The margin is also adequate to detect a single malfunctioning circulator thus avoiding ASME Code allowable steam generator tube temperature limits being exceeded. O .

                                              /                Attachment 4 P-85214 o

U ANALYSIS VALUE FUNCTION: Rod Withdrawal Prohibit FUNCTIONAL UNIT: STARTUP Channel - Low Count Rate ANALYSIS VALUE: 2.5 cps PRESENT TECHNICAL SPECIFICATION TRIP SETTING: > 2.5 cps PRESENT LICENSING BASIS: Trip is set to prevent rod withdrawal and reactor start-up without adequate neutron flux indication. The trip level is selected to be above the background noise level. The neutron startup sources are sized to produce a count rate of > 5 cps at the detectors after a plant shutdown lasting two months-- (FSAR Secticn 7.3.1.1). SAFETY LIMIT: None. Analysis of a rod withdrawal accidents at source power with the failure of low-count-rate RWP and subsequent rod withdrawal prohibits with eventual scram at 140% power results in bm no expected fuel particle failures. (FSAR Section 14.2.2.7). LIMITING INITIATING EVENT: Rod withdrawal from source power with the core at room temperature with partial coolant flow. TRANSIENT 0VERSHOOT: N/A. The count rate represents an essentially steady-state core condition which must exceed the trip-set value in order to release the prohibit. TRANSIENT TIME RESPONSE: N/A. The count rate represents an essentially steady-state core condition. PROCESS MEASUREMENT ACCURACY: Not applicable. DISCUSSION: During a rise-to-power from a shutdown condition, the operator places the Interlock Sequence Switch in the STARTUP position. this activates startup channels 1 and 2 which monitor neutron flux at very low power levels. If there is an inadequate flux reading (< 2.5 cps), a rod withdrawal prohibit occurs, preventing startuli. (Ref: P-80340). It is expected that the neutron flux reading will normally be well above the trip set point, but sin'ce the activity level of the neutron sources varies with time, depending both upon decay and irradiation O Attachment 4 P-85214 O ANALYSIS VALUE times, the detector readings also vary with time. Data obtained during reactor refueling on February 12, 1984, when the fuel element containing the neutron source was inserted into the core gave i readings of approximately 70 cps on the furthest detector.

                .(Ref: P-84081,03/13/84 transmittingLER84-003) i k

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Attachment 4 P-85214 { ANALYSIS VALUE FUNCTION: Rod Withdrawal Prohibit FUNCTIONAL UNIT: Linear Channels - 5% and 30% RWP Channels 3, 4 and 5, and 6, 7, and 8 ANALYSIS VALUE: Not Applicable. These Functional Units are Administrative Requirements only. PRESENT TECHNICAL SPECIFICATION TRIP SETTING: < 5% and < 30% LICENSING BASIS: Power indication from these channels assures that power is not increased above these levels without advancing the interlock ~ sequence switch to the appropriate position " Low Power" or

      " Power".

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Attachment 4 P-85214 ANALYSIS VALUE FUNCTION: Rod Withdrawal Prohibit FUNCTIONAL UNIT: Startup Channel - Low Count Rate ANALYSIS VALUE: 2.5 cps PRESENT TECHNICAL SPECIFICATION TRIP SETTING: > 2.5 cps LICENSING BASIS: Trip is set to prevent rod withdrawal and reactor startup without adequate neutron flux indication. The trip level is selected to be above the background noise level. The neutron startup sources are sized to produce a count rate of > 5 cps at the detectors after a plant shutdown lasting two months. (TSAR Section 7.3.1.1) SAFETY LIMIT: None. Analysis of a rod withdrawal accident at source power with the failure of low-count-rate RWP and subsequent rod withdrawal prohibits and with eventual scram at 140% power results in p no expected fuel particle failures. (FSAR Section 14.2.2.7) v LIMITING INITIATING EVENT: Rod withdrawal from source power with the core at room temperature with partial coolant flow. TRANSIENT OVERSHOOT: Not Applicable. The count rate represents an essentially steady-state core condition which must exceed the trip-set value in order to release the prohibit. TRANSIENT TIME RESPONSE: Not Applicable. The count rate represents an essentially steady-state core condition. PROCESS MEASUREMENT ACCURACY: Not Applicable. DISCUSSION: During a rise-to-power from a shutdown condition, the operator places the Interlock Sequence Switch in the STARTUP position. This activates startup channels 1 and 2 which monitor neutron flux at very low power levels. If there is an inadequate flux reading (< 2.5 cps), a rod withdrawal prohibit occurs, preventing startup. (Ref: P-80340) It is expected that the neutron flux reading will nonnally be well above the trip set point, but since the activity level of the neutron sources varies with time, depending both upon decay and irradiation times, the detector readings also vary with time. Data obtained O

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

4 Attachment 4 P-85214 O ANALYSIS VALUE 4 , during reactor refueling on February 12, 1984, when the element block i' containing the neutron source was inserted into the core gave readings of approximately 104 cps on the nearest startup detector and J approximately 70 cps on the furthest detector. (Ref: P-84081,3/13/84 transmitting LER 84-003) f lO I i O I L.

Attachment 4 P-85214 O ANALYSIS VALUE FUNCTION: Rod Withdrawal Prohibit FUNCTIONAL UNIT: STARTUP Channel and Wide Rar.ge Channel - High Rate of Change: Channels 1, 2 and Channels 3, 4, and 5 ANALYSIS VALUE: 2decadesperminute(DPM) PRESENT TECHNICAL SPECIFICATION TRIP SETTING: Currently not contained in the Technical Specifications. LICENSING BASIS: High rate neutron flux rise is used both as a rod withdrawal prohibit and a reactor scram input when the Interlock Sequence Switch is in the STARTUP position. The rod withdrawal analysis value was selected to be above nonnal operating levels to preclude spurious rod withdrawal prohibits. FSAR Table 7.1-3 states the normal operating value is less than 2 decades per minute. SAFETY LIMIT: None. FSAR Section 14.2.2.7 assumed the rod b3 withdrawal accident from source power was terminated by the 140% of rated power flux scram. In this case, there was no fuel failure or breach of the primary coolant pressure boundary. For scram to occur at 140% of rated power, the 2 DPM rod withdrawal and the 5 DPM scram would have to fail, in addition to the failure of the 5%, 30% and 120% of rated power rod withdrawal prohibits. LIMITING INITIATING EVENT: Control rod pair accidental withdrawal at less than 5% power (FSAR Section 14.2.2.7). TRANSIENT OVERSHOOT: Control rod withdrawal accident from source power terminated by a 2 DPM rod withdrawal prohibit was not specifically analyzed since analysis of a more limiting event resulted in acceptable consequences. TRANSIENT TIME RESPONSE: Not analyzed as stated under " Transient Overshoot." PROCESS MEASUREMENT ACCURACY: Not Applicable. DISCUSSION: The actual trip setting utilized for the 2 DPM rod withdrawal prohibit has been 1.5 DPM. This trip setting has not been a hindrance to plant operations in the form of unnecessary spurious rod withdrawal prohibits during plant startup operations. No credit Attachment 4 P-85214 O ANALYSIS VALUE was necessary or assumed in the FSAR for any control rod withdrawal accident being terminated by this rod withdrawal prohibit. It is therefore concluded based upon prior acceptable plant startup operations that the high rate of neutron flux rise rod withdrawal prohibit trip setting should continue to be 1.5 DPM. Further, the allowable value as defined by the ISA Standard 67.04-1982 should be "Not Applicable" since credit it not assumed in the FSAR analysis of control rod withdrawal accidents from source power being terminated by a rod withdrawal prohibit. 4 O c O

Attachment 4 P-85214 O ANALYSIS VALUE FUNCTION: Rod Withdrawal Prohibit FUNCTIONAL ' UNIT: Neutron Flux Level - Channels 3, 4 and 5 or 6, 7, and 8 ANALYSIS VALUE: 120% of rated power PRESENT- TECHNICAL SPECIFICATION TRIP SETTING: Currently not contained in the Technical Specifications LICENSING BASIS: The 120% of rated power control prohibit is one of the defenses in depth.for reactivity excursions including control rod withdrawal accidents. The analysis value was set half way between full power operation and the 140% of rated power flux scram. This is the same as used at Peach Bottom. The FSAR (Section 14.2.2.6 and

   -14.2.2.7) did not assume that a control rod withdrawal accident was
   . terminated by any of the control rod withdrawal prohibits including

- the 120% of rated rod withdrawal prohibit. SAFETY LIMIT: None. FSAR Sections 14.2.2.6 and 14.2.2.7 assume rod withdrawal accidents are not terminated by rod withdrawal prohibits but by the 140% of rated flux scram or some other later action. When terminated by the 140% flux scram from either source power or full power operation, there is not fuel failure or breach of the primary coolant boundary. If the. rod withdrawal prohibit at 120% rated power limits the reactivity excursion, system temperatures changes would be less (Ref: FSAR Section 14.2.2.6). LIMITING INITIATING EVENT: Control rod pair accidental withdrawal from 100%. rated power (F5AR Section 14.2.2.6). TRANSIENT OVERSHOOT: Control rod withdrawal accident terminated by a rod withdrawal prohibit was not specifically analyzed. TRANSIENT TIME RESPONSE: Control rod with.irawal accidents terminated

   .by a rod withdrawal prohibit was not specifically analyzed.

PROCESS MEASUREMENT ACCURACY: Control rod shim bank movement can result in decalibration of the ex-core flux monitors. The rod

   . withdrawal prohibit trip settings are administratively set based upon indicated power at less than 120% rated power to account for this
   - potential decalibration. The figure from GA Document No. 907177A.

O

Attachment 4 P-85214

 ]

ANALYSIS VALUE transmitted by GP-2191 dated 2/14/84 are both the trip settings for the scram and the rod withdrawal prohibit. The figure is contained in Linear Channel-High _ Scram Analysis Value package. The trip settings shown include allowance for ex-core detector decalibrations as well as instrumentation inaccuracy. DISCUSSION: The linear channel 120% rated power rod withdrawal prohibit is the first line of defense for rod withdrawal accidents occurring when the Sequence Selector Switch is in the POWER position (> 30% power). For control rod withdrawal accidents, it is backed by the 140% rated power scram, operator manual scram and finally by reheat steam temperature - high scram. O ! O l L

   .....:i-O I

P-85214 ATTACHMENT 5 Detector Decalibration Curves O 4 J

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Attachment 5 P-85214 Proper Implementation of Detector Decalibration Curves The Detector Decalibration Curves approved by the NFSC require b3 regular calibration of the detectors. The detectors are to be calibrated: at least once every 24 hours while in the POWER, LOW POWER, or STARTUP MODES; after each control rod group (i.e. groups 2B through 3D) is fully withdrawn and/or fully inserted; when control rod groups 3B and 3D are one-half withdrawn; whenever any channel approaches an RWP TRIP SETP0 INT; when the Indicated Thermal Power is between 2% and 4% of Rated Thermal Power; when increasing power, the ISS in the POWER MODE, and the heat balance power is between 24% and 28% of RATED THERMAL POWER; when decreasing power, the ISS is in the POWER MODE, and the heat balance goes below 36% of RATED THERMAL POWER. The detectors are to be calibrated anytime the operator has reason to believe that one or more detectors are giving anomalous readings or when individual detectors differ by more than 10%. O O - -

Attachment 5 P-85214 O

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         .O P-85214 ATTACHMENT 6 Summary of Proposed Changes O

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Attachment 6 P-85214 Summary of Proposed Changes to Fort St. Vrain Technical Specifications Section 2.0 " Definitions" Page 2-la I. Add definition for ACTU/. TION LOGIC TEST. II. Add definition for CHANNEL FUNCTIONAL TEST. Section 5.0 " Surveillance Requirements" Pages 5.0-1, -I. Replace these pages with new pages 5.0-2, and formatted after the model Westinghouse 5.0-3 Standard Technical Specifications. Specification 5.0.6 includes provisions for entry into an OPERATIONAL MODE with exceptions. Specification 5.0.8 is added to extend the surveillance intervals in certain instances. Page 5.4-1 I. This page replaces old pages 5.4-1 and 5.4-2 to incorporate the format of the model Westinghouse Standard Technical

 !],_s                                     Specifications.

Pages 5.4-2 and I. These two pages reformat Table 5.4-1 to 5.4-2a the model Westinghouse Standard Technical Specification format with few changes to the surveillances. Pages 5.4-2b and I. These are new pages and . includes 5.4-2c notes to more fully describe the surveillances being performed as a result of Table 5.4-1. Few changes have been made to the present FSV Technical Specifications. II. Note (t) has been revised to include only the six PPS temperature indicators instead of all eight indicators. l Pages 5.4-3, I. These three pages reformat Table 5.4-2 5.4-3a and to the model Westinghouse Standard 5.4-3b Technical Specification format with r few changes to the~ surveillances. I Page 5.4-3c I. This is a new page and includes notes to more fully describe the surveillances being performed as a O result of Table 5.4-2. Few changes V have been made to the present FSV l Technical Specifications. l [

1 Attachment 6 P-85214 Pages 5.4-4 and I. These two pages reformat Table 5.4-3 to 5.4-4a the model Westinghouse Standard

 ~O                        Technical Specification format with few changes to the surveillances.

Page 5.4-4b I. This is a new page and includes notes to more fully describe the surveillance being performed as a result of Table 5.4-3. Few changes have been made to the present FSV Technical Specifications. Pages 5.4-5 and I. These two pages reformat Table 5.4-4 5.4-Sa to the model Westinghouse Standard Technical Specification format and add Functional Units not included in the l present FSV Technical Specifications. Page 5.4-5b I. This is a new page and includes notes to more fully describe the surveillances being performed as a result of Table 5.4-4. Page 5.4-6 I. This is a new Table 5.4-5 and includes the surveillances required for the undervoltage protection. (O Page 5.4-6a I. This is a new page and includes notes to more fully describe the surveillances being performed as a result of Table 5.4-5. Page 5.4-7 I. This is a new page to introduce the Basis for Specification SR 5.4.1. Page 5.4-8 I. This is a new page indicating the Basis ! for SR 5.4.1. This information is included in the present format on pages 5.4-1 and 5.4-2. ( II. The description of the Basis has been ! revised slightly for this edition. Pages 5.4-9 through 5.4-11 have been eliminated with this revision. l l t l O i

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e O . , l l P-85214 ATTACHMENT 7 Proposed Changes O ,

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l i l Fort St. Vrain #1 1 i O Technical Specifications V Amendment No. Page 2-la 2.lf TOTAL NO. OF CHANNELS The TOTAL NO. OF CHANNELS shall designate the sum of channels installed to provide a trip signal to .a protective device. 2.lg TRIP SETPOINT The TRIP SETPOINT is the least conservative "as left" value (as indicated) on a protective device to prevent a measured quantity from exceeding the ALLOWABLE VALUE. l 2.1h ACTUATION LOGIC TEST An ACTUATION LOGIC TEST shall be the application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output. , l 2.11 CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of O a simulated signal into the channel as close to the sensor as practicable, considering system design, to verify that alanns, interlocks, and/or trip functions are OPERABLE. O

Fort St. Vrain #1 Technical Specifications ([]) Amendment No. Page 5.0-1 l5.0 SURVEILLANCE REQUIREMENTS l 5.0.1 The surveillance requirements specified in this section define the tests, calibrations, and inspections which are necessary to verify the performance and operability of equipmant essential to safety during designated OPERATIONAL MODES, or required to prevent or mitigate the consequences of abnormal situations. 5.0.2 Surveillance Requirements shall be applicable during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement. , l 5.0.3 SURVEILLANCE FRE00ENCIES Where surveillance frequencies or FREQUENCY NOTATIONS are identified, surveillances shall be performed at least ' once per SURVEILLANCE INTERVAL as specified " below. Any surveillance frequency preceded by at least" shall be considered a nominal time interval to f-'g which the extension permitted by Specification 5.0.4 may be applied. Other intervals may be specified in (_/ the Surveillance Requirements. SURVEILLANCE r

 .                                 FREQUENCY NOTATION                   INTERVAL f
                                          ~

i {, S 12 hours

• l D 24 hours **

l W 7 days *** e l l

• Two 12 hour SURVEILLANCE INTERVALS shall require performance of the specified surveillances once during any of the AM hours and once,during.any of the PM hours.

l . l

                      ** A 24 . hour SURVEILLANCE INTERVAL shall require performance of the specificd surveillances once per calendar day.                ,
                    *** A 7 day SURVEILLANCE INTERVAL shall require performance of the specified survel.11ances once per calendar week.              .

l O

                                                                                              --__----._-----.u.     - - _

Fort St. Vrain #1 Technical Specifications

    ,3 Q                                                             Amendment No.

Page 5.0-2 FREQUENCY SURVEILLANCE NOTATION INTERVAL l M 31 days l Q 92 days l SA 184 days l A 366 days l R Refueling Cycle P Prior to each reactor startup, if not performed within previous 7 days 5.0.4 Each Surveillance Requirement shall be performed within the specified time interval with:

a. A maximum allowable extension not to exceed 25% of the surveillance interval, and O b. The combiaed time intervei for eny a ceasecutive surveillance intervals not to exceed 3.25 times the specified SURVEILLANCE INTERVAL.

5.0.5 Failure to perform a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications. Surveillance Requirements do not have to be performed on inoperable equipment. 5.0.6 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation have been performed within the stated surveillance interval except surveillances that cannot be performed prior to entry into an OPERATIONAL MODE shall be performed within 72 hours after entry into that MODE. l 5.0.7 Implementation of the in-service inspection (ISI) surveillance requirements shall be per one of the following criteria, unless otherwise indicated: ISI Criterion A: The surveillance requirement shall h be implemented before 90 days have elapsed following the formal

Fort St. Vrain #1 n Technical Specifications U Amendment No. Page 5.0-3 approval date of Amendment No. 33 by the Commission. ISI Criterion B: The surveillance requirement shall be implemented before the beginning of fuel cycle 4, provided that fuel cycle 4 does not begin within 90 days from the formal approval date of Amendment No. 33 by the Commission. Othemi se, the surveillance requirement shall be implemented before the end of the first scheduled plant shutdown followirg 90 days from the formal approval date of Amendment No. 33 by the Commission. ISI Criterion C: The surveillance requirement shall be implemented before the beginning ' of fuel cycle 5. O ISI Criterion D: The surveillance requirement shall be implemented in the existing l schedule of surveillance tests, following 90 days from the formal l approval date of Amendment No. 33 by the Commission. I 1 l t I lO j__, . . , - . . . _ . . . . . ,

Fort St. Vrain #1 ( Technical Specifications l Amendment Page 5.4-1 INSTRUMENTATION AND CONTROL SYSTEMS O ll 5.4 PLANT PROTECTIVE SYSTEM INSTRUMENTATION SURVEILLANCE AND CALIBRATION REQUIREMENTS 5.4.1 Each PLANT PROTECTIVE SYSTEM channel and interlock and the automatic trip logic and the 480V essential bus undervoltage protection shall be demonstrated OPERABLE by the performance of the PLANT PROTECTIVE SYSTEM Instrumentation Surveillance and Calibration Requirements specified in Tables 5.4-1 through 5.4-5 and associated notes. C' O

Fort St. Vrain #1 Technical Specifications Amendment Page 5.4-2 O i ' Saecificatioa SR s.4.1 l Table 5.4-1 l REACTOR SCRAM SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL ACTUATION FUNCTIONAL CHANNEL CALIBRA- FUNCTIONAL LOGIC APPLICABLE NO. UNIT CHECK TION TEST TEST MODES ! la. Manual Scram P(aa) P,L,S/U S/0,R(a}

l. (Control Room)

P.M(ab) P lb. Manual Scram R(LlS/U, a (Outside Control Room) 2a. STARTUP D R(b) P(c) R(a) Channel-High 2b. Wide D M(d),R(e) P(c) S/U Range Channel Rate of Change-High 3a. Linear D D(d),R(e) M(c) P,L,S/U O' Channel-High Channels 3,4,5 3b. Linear D 0(d),R(e) M(c) P,L,S/U Channel-High Channels 6,7,8

4. Primary Coolant Moisture a) High Level D(ac),R(ae) M(af) M(ag) P.L Monitor D(ad) b) Loop Monitor c , R(ae) ) M(ai)

5. Reheat D(aj) R(f)(e) M(ak) P,L Steam Temperature

                          -High P

6a. Primary D(al) R(am)(e) M(ak) Coolant Pressure

                          -Programed Low O

Fort St. Vrain #1 Technical Specifications Amendment Page 5.4-2a l Specifi ation SR 5.4.1

             ]                              Table 5.4-1 (Continued) l         REACTOR SCRAM SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL     CHANNEL      ACTUATION FUNCTIONAL       CHANNEL   CALIBRA- FUNCTIONAL        LOGIC    APPLICABLE NO.      UNIT          CHECK       TION      TEST          TEST            MODES 6b. Circulator           0(an) R(f)(e)      M(ak)                   PL Inlet Temperature for Programed Pressure

7. Primary D(al) R(am)(e) M(ak) M(ao) PL Coolant Pressure

                       -Programed High

8. Hot Reheat R(g) M(ap) P Header Pressure

                       -Low O                9. Main Ste-Pressure Rcs)       McaP)                   e
                       -Low

10. Plant R(au). M(aq) P.L.S/U Electrical System-Loss

11. Two Loop M(ar) R(as) P,L,S/U Trouble Scram Logic

12. High Reactor D R(f) M(ak) P,L,S/U Building Temperature (Pipe Cavity)

_ . . . _ _ w _ . _ Fort St. Vrain #1-Technical Specifications Amendment Page 5.4-2b O i Specificatioa SR 5.4.1 l Table 5.4-1 (Continued) l TABLE NOTATION (a) The applicability for the REFUELING MODE is anytime the reactor scram is reset and the control rod drive system is capable of rod withdrawal. (b) Internal test signal shall be checked and calibrated to assure that its output is in accordance with the design requirements. This shall be done after completing the external test signal procedure by checking the output indication when turning the internal test signal switch. (c) An internal test signal is applied to verify TRIP SETPOINTS and alarms. (d) The channel is adjusted to agree with the heat balance calculation when the reactor is critical. (e) An internal test signal is applied to adjust TRIP SETPOINTS and indicators as required. l(f) Compare each thermocouple output to an NBS traceable standard. l(g) Known pressure applied at sensor to adjust TRIP SETPOINTS. (aa) Manually trip the system. Verify trips, alarms, and indications including proper operation of scram contactors. l(ab) Manually trip each channel. (ac) Verification of sample flow of OPERABLE moisture monitors per note (f) of Table Notation for Table 4.4-1. l(ad) Compare two mirror temperature indications. l(ae) Inject moisture laden gas into sample lines. I (af) Verify that each of the OPERABLE moisture monitors sample flow alarms are OPERABLE. (ag) Trip one. high level and one low level channel, pulse another low level channel. l(ah) Trip each channel and verify proper indications. (ai) Trip each channel, pulse test other loop to check loop identification. l(aj) Compare the averaged themocouple channel input indications.

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Fort St. Vrain #1 Technical Specifications Amendment Page 5.4-2c

 .O         cax) Trip chaaaei . verify alar =s aad iadicatieas- rateraai test
                  . signal to verify TRIP SETPOINTS and alarms.

l(al) Comparison of three programmed channel indications. l(am) Known pressure applied to sensor. l(an) Comparison of six separate PPS temperature indications. (ao) Pulse test one channel with another channel tripped and verify proper indications, both channels. (ap) Reduce pressure at sensor to trip channel; verify alarms and indications. (aq) Trip each channel by applying simulated loss of voltage signal, verify alarms and indications. (ar) Special test module used to trip channel by energizing each of four appropriate pairs of two-loop trouble relays. l(as)'Triplogictocausetwolooptroublescram. l(at) Permissible bypass conditions l 1. Any circulator buffer seal malfunction l 2. Either loop hot reheat header high activity l 3. As stated in LCO 4.9.2 (au) Trip each relay by applying simulated loss of voltage signal; adjust TRIP SETPOINTS and time delays and verify alarms and indications. j l(av) Compare six mirror temperature indications. I l O l l l

Fort St. Vrain #1 Technical Specifications Amendment Page 5.4-3 Oi Saeciricatioa SR S.4.1 l. Table 5.4-2 l LOOP SHUTDOWN SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL ACTUATION FUNCTIONAL CHANNEL CALIBRA- FUNCTIONAL LOGIC APPLICABLE NO. UNIT CHECK TION TEST TEST MODES la. Steam Pipe D R(ba) M(bb)(bc)M(bd)(be) P,L,S/U Rupture Under PCRV, Loop 1 lb. Steam Pipe D R(ba) M(bb)(bc) M(bd)(be) P,L,S/U Rupture Under PCRV, Loop 2 1c. Steam Pipe D R(ba) M(bb)(bc) M(bd)(be) P,L,S/U Rupture, North Pipe Cavity Loop 1 Id. Steam Pipe D R(ba) M(bb)(bc) M(bd)(be) P,L S/U Rupture, South Pipe Cavity O ' a' le. Steam Pipe M(bb)(bc)M(bd)(be) P,L,S/U D R(ba) Rupture, North ' Pipe Cavity Loop 2 If. Steam Pipe D R(ba) M(bb)(bc) M(bd)(be) P,L,S/U Rupture, South Pipe Cavity Loop 2 2a. High Pressure, R(g) M(h) P,L,S/U Pipe Cavity R(bg) M(bh) P,L,S/U 2b. High Temperature, Pipe Cavity

                              ~

R(g) M(h) P,L,S/U 2c. High Pressure, Under PCRV R(bg) M(bh) P,L,S/U 2d. High Temperature, Under PCRV O

   - . , . .                z.      -.

Fort St. Vrain #1 Technical Specifications Amendment Page 5.4-3a O i- Spec 4ricat4oa sa s 4 1 l Table 5.4-2-(Continued) l LOOP SHUTDOWN SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL ACTUATION FUNCTIONAL CHANNEL CALIBRA- FUNCTIONAL LOGIC APPLICABLE NO. UNIT CHECK TION TEST TEST MODES 3a. Loop 1 R(bi) P,L,S/U, Shutdown (bj) Logic 3b. Loop 2 R(bi) P,L,S/U, Shutdown (bj) Logic 4a. Circulator. 1A- M(bk) P,L,S/0, and 1B Shutdown (bj)

                         -Loop Shutdown Logic 4b. Circulator 1C                                      M(bk)     P,L,S/U, and 10 Shutdown                                             (bj)

(qj -Loop Shutdown Logic Sa. Steam' Generator R(g) M(h) M(b1) P,L,S/U Penetration Overpressure Loop 1 Sb. Steam Generator R(g) M(h) M(b1) P,L,S/U Penetration Overpressure Loop 2 6a. High Reheat D(bm) R(bn)(e) M(b1) P,L,S/U Header Activity, Loop 1 6b. High Reheat D(bm) R(bn)(e) M(b1) P,L,S/U Header Activity, Loop 2 7a. Low Superheat 0(bm) R(f)(e) M(bo) P Header Temperature, Loop 1 O

Fort St. Vrain #1 Technical Specifications Amendment Page 5.4-3b l Specification SR 5.4.1 O U Table 5.4-2'(Continued) l l LOOP SHUTDOWN SYSTEM INSTR MENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL ACTUATION FUNCTIONAL CHANNEL CALIBRA- FUNCTIONAL LOGIC APPLICABLE NO. UNIT CHECK TION TEST TEST MODES 7b. Low Superheat O(bm) R(f)(e) M(bo) P Header Temperature. Loop 2 7c. High D(bm) R(f)(e) M(bo) P Differential Temperature Between loop 1 and Loop 2

8. Primary Coolant-Moisture a) High Level ---------------See Table 5.4-1------------------

l b) Loop Monitor ---------------See Table 5.4-1------------------ O O

_. ., _ ._ a _ _ __. _ . _ _ _ . _ _ _ Fort St. Vrain #1 Technical Specifications i Amendment Page 5.4-3c Ol Specification SR 5.4.1

         'l                              Table 5.4-2 (Continued)

TABLE NOTATION l l(d) Compare each thennoccuple output to an NBS traceable standard.

              '(e)    An internal test signal is applied to adjust TRIP SETPOINTS and indicators.

l(f) Compare each thermocouple output to an NBS traceable standard. l(g) Known pressure applied at sensor to adjust TRIP SETPOINTS. l(h) Pressure switch actuated by pressure applied at sensor. l(ba) Known sound reference applied at sensor to adjust channel gain. [(bb) Internal signal to adjust ultrasonic TRIP 3ETPOINTS. l(bc) Trip test signal solenoid valves to verify loop integrity. (bd)-Pulse test one temperature channel with another temperature channel tripped, while simultaneously having two ultrasonic channels tripped. (be) Pulse test one pressure channel with another pressure channel tripped, while simultaneously having two ultrasonic channels tripped. l(bg) Known temperature applied at sensor to adjust TRIP SETPOINTS. l(bh) Temperature switch actuated by heat applied at sensor. _[(bi) Trip both circulators to test loop shutdown. (bj) Applicable logic for each MODE depends upon the applicable Functional Unit parameters for the MODE and which circulator, circulator drive, and loop are operating or are required to be OPERABLE. l(bk) Pulse test and verify proper indications. (bl) Pulse test each channel with another channel tripped and verify proper indications. [(bm) Comparison of three separate indications in each loop. l(bn) Expose sensor to known radiation source for calibration. (bo) Pulse test one temperature channel and one temperature differential channel with one temperature and one temperature O differential channel tripped.

Fort St. Vrain #1 Technical Sp;cifications Amendment Page 5.4-4 l Specification SR 5.4.1 U l Table 5.4-3 l CIRCULATOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL ACTUATION FUNCTIONAL CHANNEL CALIBRA- FUNCTIONAL LOGIC APPLICABLE NO. UNIT CHECK TION TEST TEST MODES

1. Manual Trip R(ca) All Modes (Steam) (cb)

2. Circulator D R(cd) M(ce) M(cf) All Modes Speed-High (cb)

(Steam)

3. Circulator D R(g) M(cf) All Modes Drain Mal- (cg) function

4. Manual Trip R(ch) All Modes (Water) (cb)

5. Circulator D R(cd) M(ce) M(cf) All Modes Speed-High (cb)

(Water) O e. Circuiator Speed-Low D R(cd) Mcce) Mccf) e Programed 7a. Loop 1, D R(e)(ci) M(ce) M(cf) P Fixed Feedwater Flow-Low (Both Circulators) 7b. Loop 2 D R(e)(ci) M(ce) M(cf) P Fixed Feedwater Flow-Low (Both Circulators) 8a. Loop 1, , D R(e)(ci) M(ce) M(cf) P Programed Feedwater Flow-Low (Both Circulators) 4

                                         , - -

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Fort St. Vrain #1 Technical Specifications Amendment Page 5.4 da l Specification SR 5.4.1 Table 5.4-3 (Continued) CIRCULATOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS CHANNEL CHANNEL ACTUATION FUNCTIONAL CHANNEL CALIBRA- FUNCTIONAL LOGIC APPLICABLE NO. UNIT CHECK TION TEST TEST MODES 8b. Loop 1, D R(e)(ci) M(ce) M(cf) P Programed Feedwater Flow-Low (One Circulator) 9a. Loop 2, D R(e)(ci) M(ce) M(cf) P Programed Feedwater Flow-Low (Both Circulators) 9b. Loop 2, D R(e)(ci) M(ce) M(cf) P Programed Feedwater Flow-Low (One Circulator) O to circui tor Seal o a(e)(ca) acce) acc<) ^11 aodes (cb) Malfuntion-Low . 10b. Circulator D R(e)(cj) M(ce) M(cf) All Modes Seal (cb) j Malfunction-High

11. Loss of D R(g) M(cf) All Modes Circulator (cb)

Bearing Water

12. Circulator R(g) M(h) M(cf) All Modes Penetration (cb)

Overpressure l I 1 I l l O 1 i

Fort St. Vrain #1 Technical Specifications Amendment Page 5.4-4b Specification SR 5.4.1 O ll Table 5.4-3 (Continued) l TABLE NOTATION (e) An internal test signal is applied to adjust TRIP SETPOINTS and indicators. l(g) Known pressure applied at sensor to adjust TRIP SETPOINTS. l(h) Pressure switch actuated by pressure applied at sensor. (ca) Trip steam turbine drives. Verify water turbine automatic start. (cb) Applicable in MODES where the circulator is required to be OPERABLE per LCO 4.2.1 and LCO 4.2.2. s (cd) A known pulse frequency is applied to adjust TRIP SETPOINTS and indicators. (ce) An internal test signal is applied to verify TRIP SETPOINTS and indicators. (cf) Pulse test one channel with another channel tripped and verify proper indications. in the SHUTDOWN MODE is when the helium (cg) Applicabilitfisbeingdrivenbysteam. circulator (s l(ch) Trip water turbine drives and verify proper indications. (ci) A known differential pressure is applied at the flow transmitter. l(cj) A known differential pressure is applied at the sensor. l(ck) Pressure switch actuated by pressure applied at sensor. O

w -~u. - .-.- . - - . .-. . - -. _ - .. ... - . - _ . - _ - Fort St. Vrain #1 Technical Specifications Amendment Page 5.4-5 l Specification SR 5.4.1 l . Table 5.4-4 l ROD WITHDRAWAL PROHIBIT INSTRUMENTATION SURVEILLANCE RECUIREMENTS CHANNEL CHANNEL ACTUATION FUNCTIONAL CHANNEL CALIBRA- FUNCTIONAL LOGIC. APPLICABLE NO. UNIT CHECK TION TEST TEST MODES

1. STARTUP D R(b) P(c) S/U R(b)

Channel-Low (da) Count Rate (Channels 1 and 2) 2a. Linear D D(d), M(c) S/U(db) Channel R(e) 5% RWP (Channels 3,4 and5) 2b. Linear D D(d), M(c) S/U(db) Channel R(e) 5% RWP (Channels 6,7 and8) O 3a. Linear D 0(d), M(c) L(de) Channel R(e) 30% RWP - (Channels 3,4 and 5) 3b. Linear D 0(d), M(c) L(de) Channel R(e) 30% RWP (Channels 6,7 and 8) 4a. Wide Range D M(d), P(c) S/U Channel Rate R(e) of Change-High (Channels 3,4 and 5) 4b. STARTUP D R(b) P(c) S/U(da) Channels Rate of Change-High (Channels 1 and 2) O

Fort St. Vrain #1 Technical Specifications Amendment Page 5.4-Sa Oi Saeciricetioa SR 5.4.1 l Table 5.4-4 (Continued) l R0D WITHDRAWAL PR0HIBIT INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL ACTUATION FUNCTIONAL CHANNEL CALIBRA- FUNCTIONAL LOGIC APPLICABLE N0. UNIT CHECK TION TEST TEST MODES Sa. Linear D D(d), M(c) P.L.S/U Channel-High R(e) Power RWP (Channels 3,4 and 5) Sb. Linear D D(d), M(c) P,L,S/U Channel-High R(e) Power RWP (Channels 6, 7 and 8)

6. Multiple P(dd) All Modes Rod Pair R(de) (a)

Withdrawal O O

Fort St. Vrain #1 Technical Specifications Amendment Page 5.4-5b q l Specification SR 5.4.1 V Table 5.4-4 (Continued) l TABLE NOTATION l (a) The applicability for the REFUELING MODE is anytime the reactor scram is reset and the control rod drive system is capable of rod withdrawal. (b) Internal test signal shall be checked and calibrated to assure that its output is in accordance with the design requirements. This shall be done after completing the external test signal procedure by checking the output indication when turning the internal test signal switch. (c) An internal test signal is applied to verify TRIP SETPOINTS and alams. (d) The channel is adjusted to agree with the heat balance calculation when the reactor is critical. (e) An internal test signal is applied to adjust TRIP SETPOINTS and indicators. (da) The start-up channel may be disabled above E-03% of RATED THERMAL POWER. O C' automatically (db) The Low Power RWP bistable resets at approximately 4% of RATED THERMAL POWER after reactor power is decreased from greater than 5% of RATED THERMAL POWER. Power Range RWP bistables automatically reset at (de) The approximately 10% of RATED THERMAL POWER after reactor power is decreased from greater than 30% of RATED THERMAL POWER. The RWP may be manually reset between approximately 10% and 30% of RATED THERMAL POWER. (dd) Attempt a two rod pair withdrawal. Check for actuation of prohibit. (de) A current is simulated through the sensor to verify actuation and alarms. O

Fort St. Vrain #1 Technical Specifications Amendment Page 5.4-6 -s l Specification SR 5.4.1 (") l Table 5.4-5 l 480 VAC ESSENTIAL BUS UNDERVOLTAGE PROTECTION CHANNEL CHANNEL ACTUATION FUNCTIONAL CHANNEL CALIBRA- FUNCTIONAL LOGIC APPLICABLE NO. UNIT CHECK TION TEST TEST MODES

1. Plant --------------See Table 5.4-1--------------

Electrical System-Loss (Scram)

2. Degraded R(ea) M(eb) All Modes Voltage

3. Loss of R(ea) M(eb) All Modes Voltage-Automatic Throwover (ATO)

4. Loss of R(ea) M(eb) All Modes Voltage-D.G.

Start, Load A Shed and Load V Sequence O

Fort St. Vrain #1 Technical SpIcifications taendment Page 5.4-6a l Specification SR 5.4.1 l Table 5.4-5 (Continued) TABLE NOTATION l (ea) Trip each channel b applying a simulated voltage signal (decreased or loss of)y; Adjust TRIP SETPOINTS and time delays and verify alarms and indications. (eb) Simulate loss of voltage on one channel and verify alanns, indications, relay actuation, and time delays. O O

Fort St. Vrain di Tcchnical SpIcifications Amendment Page 5.4-7 "o's OThe Bases contained in succeeding pages sumarizes the reasons for the Specifications in Section 5.4.1, but in accordance with 10CFR50.36 are not part of these Technical Specifications. O . e 0 0 0

Fort St. Vrain il Technical Specificatior,s Amendment Page 5.4-8 l hsis For Specification SR 5.4.1

 ' The specified surveillance check and test minimum frequencies are based on established industry practice and operating experience at conventional and nuclear power plants. The testing is in accordance t;ith the IEEE Criteria (IEEE No. 279-1968) for Nuclear Power Plant Protection Systems, and in accordance with accepted industry standards.

Calibration frequency of the instrument channels listed in Tables l5.4-1,5.4-2,5.4-3,5.4-4,5.4-5 are divided into three categories: passive type indicating devices that can be compared with like units on a continuous basis; semiconductor devices and detectors that may drift or lose sensitivity; and on-off sensors which must be tripped by an external source to detemine their setpoint. Orift tests by GGA on transducers similar to the reactor pressure transducers (FSAR Section 7.3.3.2) indicate insignificant long term drift. Therefore a once per refueling cycle calibration was selected for passive devices (themocouples, pressure transducers, etc.). Devices incorporating semiconductors, particularly amplifiers, will be also calibrated on a once per refueling cycle basis, and any drift in response or bistable setpoint will be discovered from the test program. Drift of electronic apparatus is not the only consideration in determining a calibration frequency; for example, the change in power distribution and loss of detector chamber sensitivity require that the nuclear power range system be calibrated every month. On-off sensors are calibrated and tested on a once per refueling cycle basis. O . e O t

c . ;. . . .u..c -

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                                                                                                       --            -   . = -

a O 4 1 i i P-85214 , ATTACHMENT 8 Significant Hazards Considerations Analysis L i I i I l O l l.. . - . . _ _ .

m _. , Attachment 8

 $                                                                   P-85214 SIGNIFICANT HAZARDS CONSIDERATIONS ANALYSIS The proposed changes to SR 5.4.1 and Sections 2.1 and 5.0 do not revise the technical content of the Technical Specifications with the exception of the new addition of some surveillance and Table 5.4-5.

The addition of Table 5.4-5 incorporates the surveillance requirements for the Plant's Auxiliary Electrical System Undervoltage Protection. Since these proposed changes only revise the format and document the surveillances performed and do not alter the safety functions of the systems, no significant safety hazards considerations are involved. Based on the above, operation of Fort St. Vrain (FSV)in accordance with the proposed changes will not 1) involve a significant increase in the probability or consequences of an accident previously evaluated, 2) create the possibility of a new or different kind of accident from any accident previously evaluated, or

3) involve a significant reduction in a margin of safety.

9

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