ML20236B218

Forwards Rev 5 to Updated FSAR for Fort St Vrain.Rev Includes Effects of Changes Made to Facility or Procedures, Safety Evaluations & Analyses of New Safety Issues Performed by or on Behalf of Util Per NRC Request

ML20236B218
Person / Time
Site:	Fort Saint Vrain
Issue date:	07/22/1987
From:	Robert Williams PUBLIC SERVICE CO. OF COLORADO
To:	Calvo J NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation
Shared Package
ML20236B219	List: ML20236B218
References
P-87522, NUDOCS 8707290012
Download: ML20236B218 (15)

Text

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"'r t-4 0~ PublicService-  :::t: A don'vIr7cYac201 oe40 i A

For t. r n $OpESgjS, JR.

Unit No. 1 NUCLEAR OPERATIONS P-87255 U. S. Nuclear Regulatory Commission ATTN: Document Contrc1 Desk Washington, DC 20555 Attention: Mr. Jose A. Calvo Director, Project Directorate IV Docket No. 50-267

SUBJECT:

Fort St. Vrain Updated FSAR, Revision 5

REFERENCE:

1) PSC Letter Brey to Gagliardo dated July 16, 1987 (P-87246)

Dear Mr. Calvo:

Enclosed is one (1) signed original and ten (10) additional copies of Revision 5 to the Updated Final Safety Analysis Report (Updated FSAR) for Fort St. Vrain (FSV) which has been prepared and is being submitted in accordance with 10CFR50, Section 50.71(e).

Thic 4 vision includes the effects of changes made to the facility or procedures as described in the Updated FSAR; safety evaluations performed by PSC, either in support of license amendments or in support of conclusions that changes did not involve an unreviewed safety question; and analyses of new safety issues performed by.or on behalf of PSC at Commission request.

Brief descriptions of the principal changes or additions to affected sections and appendices of the Updated FSAR, Revision 5, are included in Attachment A. 10CFR50.71(e)(2)(ii) requires that this submittal identify changes made under the provisions of 10CFR50.59 but not previously submitted to the Commission.- Most changes reflected in Revision 5 of the Updated FSAR are identified in the current 10CFR50.59 annual report (Reference 1). Other changes made in the FSAR, not requiring prior Commission approval pursuant to 10CFR50.59(a), and not yet reported to the Commission, are listed in Attachment B of this letter. These changes will be more fully described along with the results of their safety evaluations, as appropriate, in PSC's next annual report of 10CFR50.59 changes.

Attachment C contains a summary of PSC's resolution of comments

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provided by the NRC on Revision 4 of the Updated FSAR.

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P-87255 . July 22, 1987 If you have any- questions or comments, please contact Mr. M. H. Holmes at (303) 480-6960.

Very truly,yours, s R. O. Williams, Jr.

Vice President, Nuclear Operations R0W/BRD:pa Attachments:

A - Description of Principal Changes Incorporated in Revision 5 of the Updated FSAR B - List of 10CFR50.59 Changes Not Previously Reported to the Commission 4

C - Resolution of NRC Comments Concerning Revision 4 of the Updated FSAR cc: Regional Administrator, Region IV Attention: Mr.'J. E. Gagliardo, Chief Reactor Projects Branch (with one copy of Revision 5 to the Updated FSAR)

Mr. R. E. Farrell Senior Resident Inspector Fort St. Vrain (with one copy of Revision 5 to the Updated FSAR) i

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION P In the Matter L

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Public-Service Company of Colorado ) Docket No. 50-267 c

Fort St. Vrain Unit No. 1 )

i AFFIDAVIT R. O.

Williams, Jr., being duly swora, deposes and says that he is'Vice duly President of Public Service Company of Colorado; that he is authorized to sign and file with the Nuclear Regulatory Commission the Revision 5 of the Updated FSAR; that he is familiar with the content thereof, and that the matters set forth therein are true and correct to the best of his knowledge, information, and belief.

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H. O. W1IIlams, JP. i Vice President, Nuclear Operations STATE OF [## tem de )

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COUNTY OF dnrrap )  !

i Subscribed and sworn Ao before me, a Notary Public on this j 49,r w day of th/u , 1987.

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Wnmw fo. (l&L/khtaw VQ530 Midnina0c%sde f/ta ut w , d M e fD22/

My comission expires Omnd 22 ,198f.

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" Attachment A

, to.P-87255 DESCRIPTION OF PRINCIPAL CHANGES INCORPORATED IN REVISION 5 0F THE UPDATED FSAR Section 1 Revised the Building 10 description as it applies to access from Building 10.to the ~ control room and the Turbine-Building. .Added a . discussion regarding the consequences of a total loss of HVAC in Building 10. Stated that Safe . Shutdown Cooling using' the emergency condensate header following a 90 minute interruption of forced circulation _(10FC) from 87.5% reactor power .will. not result in fuel damage or primary. coolant system damage.. Added a e ' substantial description of the Environmental Qualification (EQ)

Program which is based on 10CFR50.49. Added a high' energy line break (HELB) to the-list of accidents.that. require the use of.

Safe Shutdown Cooling equipment. Revised Tables 1.4-1 and 1.4-2 as follows: . a)-stated that the storage basin pump structure includes .the . circulating water make-up. pump house, b) added the helium circulator brake and seal. system, c) added the discharge ,

flow . path for the steam generator ' economizer-evaporator-superheater(EES)-to. atmosphere and the depressurization flow path to -the ~ steam / water dump tank, d) added the Steam Line Rupture Detection / Isolation System (SLRDIS), and e) added portions of the ventilation and heating systems as they apply to the standby diesel generator rooms, the firewater pump house, and the service water pump house.

Section 2 Updated this section to more accurately reflect the land use in the vicinity of FSV. Revised Figure 2.1-4 to accurately reflect all gas wells in the vicinity of FSV. Updated' Table 2.4-1 to list the current manufacturers located within a 20 mile radius of FSV.

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Attachment A

.t to;P-87255 Section 3 Supplied a justification; for the .use .6f half-length core.

reflector' blocks. Revised this section, including Table: 3.6-4,

.to-' differentiate 'between< a failure of one and'a failure.of two.

core region outlet thermocouple. Also, revised the repair time and-required' operator actions for core regionfoutlet thermocouple E l failure (s). Revised the fuel temperature ' calculations terminology in Sectiona3.6 for consistency. Added,the' control-rod drive (CRD)' watts :for normal outward ' shim. Incorporated a-description of the.. approved Preventive Maintenance Program for

.the Control: Rod Drive and' Orifice Assembly (CRD0A). Expanded the discussion. of .the.CRDOA refurbishment program. Stated that'the orifice assembly of the CRDOA will' withstand a .90 minute -IOFC

.followed by Safe Shutdown Cooling and still allow adequate core cooling without repositioning. The orifice. valve i_s not required-to operate under the 10FC condition, and is assumed to remain in its initial position throughoutthe entire cooldown transient.

Revised' Table 3.5-4 to list the core excess reactivity after Pa-233 has decayed for one week -and two weeks, at refueling temperature.

Sectio'n 4 Added a discussion on helium circulator operation without the use of buffer helium. Stated that Safe Shutdown Cooling using an EES' i section following .a 90 minute 10FC will not jeopardize the structural integrity of the steam generators 'from 105% reactor power and. referenced. applicable justifying analyses.- Stated that single loop power operation will be permitted only long enough to reduce power in an orderly manner to recover the shutdown loop or to perform an orderly shutdown. Added Figure 4.4-1 and revised the text .to .' incorporate operating limits for the helium circulators based on helium temperature and circulator speed.

Section 5 Revised the tendon' corrosion evaluation based 'on the third.

interim surveil. lance report. Included the effects on the PCRV, PCRV liner, and core support floor of a complete loss of PCRV liner cooling during the cooldown following a HELB or DBA-2.

Reflected the fact that temperature measuring devices have been installed on all CRD mechanisms. Revised Table 5.9-1 is ,

accurately reflect the design allowable and expected heat loads. '

for the thermal barrier.

Attachment A

.to P-87255 Section 6 Stated that no credit was taken for operation of the reactor plant ventilation exhaust system or the pressure. relief system following a HELB. Stated that the steam generator reheaters are.

no longer relied upon for Safe Shutdown Cooling following a 90 minute. 10FC. Modified the description of the steam / water dump system to more accurately reflect its operation.

Section 7 Modified the helium circulator trip design description to state that if one helium circulator trips, a second helium circulator in the same loop should not trip on overspeed. Also, the trip of

, a single helium circulator in a loop results in the' turbine generator load being reduced to 50%. Stated that extended single loop operation at power is not permitted. Clarified the discussion on the moisture monitoring system. Added a description of the design and operation of SLRDIS and information from the SLRDIS-Safety Analysis Report. This description includes the composite temperature profiles for the HELB scenarios. Added a discussion on Building 10 access, security, HVAC, and fire protection. Revised Figure 7.1-14 to remove the "and" gate from each channel of the undervoltage reactor scram inputs to PPS.

Section 8 Stated that the standby diesel generators are independent of each other to the extent that no single failure will interfere with the proper operation (manually as a minimum) of the redundant counterpart, nor result in paralleling of the standby diesel generator sets'. Reflected that one 230 KV transmission line from the FSV substation was redirected from the Pawnee substation to the Ft. Lupton switchyard. Clarified the load sequencing discussion associated with loss of all outside power. Clarified the effects of - a loss of a DC Bus with respect to automatic switching of control power to the essential 480 VAC switchgear and undervoltage relays. Updated the discussion on protection under degraded or loss-of-voltage conditions. Table 8.2-10 is provided to depict the undervoltage relays' use and the appropriate actions initiated corresponding to the degree of voltage degradation. Clarified the description of the power supplies to the control room normal and emergency lighting.

Defined essential electrical cables. Added certain exceptions to the electrical wire size requirements. Revised Figure 8.2-9 to reflect that the Bearing Water Pumps meet the 20 foot cable separation requirement for fire protection.

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t Attachment A to P-87255 Section 9-Stated that the design of the fuel storage wells will preclude criticality under dry, flooded, and partially flooded conditions.

Clarified the minimum operating conditions for the fuel storage wells to preclude any leakage of helium. Revised the discussion for the helium purification system to accurately reflect which components must be available for primary plant depressurization as required by Technical Specification Limiting Condition for Operation (LCO) 4.2.18. Clarified the description of the nitrogen recondensed chiller units and use of these chiller units as an alternate cooling source to cool the helium purification coolers. Revised the discussion of Technical Specification LC0 4.2.4 for the service water system to accurately reflect which components must be cooled by service water. Stated that the hydraulic oil system has been environmentally qualified and .will not be adversely effected by a HELB. Stated that RTV foam and cerafiber are used to seal electrical conduit ends that drop into selected cable trays to preclude water ingress from sources such as the "G" and "J" wall firewater spray system. Added that protective measures were implemented to preclude damaging components resulting from firewater spray due to a HELB. Revised Figure 9.2-2 to show a second manipulator in the Hot Service Facility. Revised Figures 9.12-1 and 9.12-2 to show the redundant firewater pump discharge cross-connect isolation

- valves. Also, revised Figure 9.12-2 to show the new firewater supply line into the emergency condensate header for Safe Shutdown Cooling.

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Attachment'A L, to P-87255

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' Section 10 Confirmed!that:the-following' condensate cooldown transients _using a steam generator EES section or reheater are acceptable . from l 83.2% reactor power: a) coincident loss'of outside electrical power and main turbine trip,_b): coincident loss of.. outside

,, electrical ' power, main turbine trip, and loss of one standby _

generator, c)' failure of condenser vacuum,. and d) simultaneous p'  : loss .of .all three boiler feed pumps. Revised the Safe. Shutdown Cooling description to include a:HELB, in addition to the ~ Design Basis Earthquake -and. Maximum Tornado, and eliminate'the use of' L steam.-generator reheaters. This' description of Safe Shutdown Cooling ..is based- on' variable - primary _ coolant helium flow to maintain subcooled-- conditions in the steam generator EES-

sections, Econsistent with-recent analyses. This also includes a description _of the new additional firewater supply line to the ,

emergency condensate header and-the new main steam six inch vent i lines to discharge secondary' coolant _ exiting the EES ~section to atmosphere. ' Stated that a single active fail'ure_'is required to be assumed following a HELB. Revised.the Safe Shutdown Cooling description following a Design Basis Earthquake or Maximum Tornado to state that a passive failure is not required to be postulated during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Stated that the boiler feedwater system is adequate to sustain the fluid. transient effects of a simultaneous closure _ of the isolation and flow valves on both loops due to SLRDIS actuation. Revised Figures 10.1 and 10.1-2 to show the new firewater supply line into the emergency condensate header, the six inch vents from the EES i sections for Safe. Shutdown _ Cooling, and the check valve installed l

-in the emergency condensate header to ensure the integrity of the l emergency condensate header when using this path' for Safe Shutdown Cooling following a HELB.

Section 11 Clarified the functions of the radioactive gas waste system ,

radiation monitors to state that the gas waste blowers are not l' automatically tripped. .Added a discussion on post-accident sampling using Radiological Emergency Response Plan (RERP) procedure RERP-CORE.

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Attachment A to P-87255 l Section 12 1 Removed Section 12.7

• References for Section 12" and placed a reference section at the end of Sections 12.1, 12.2, 12.3, 12.4, 12.5, and 12.6. Revised Section 12 based on. current practices and procedures for RERP. Updated the organizational positions, l

. responsibilities, and authorities. Revised the training and -i background of key personnel to not include personal qualifications and background of the current incumbent personnel.

This qualification and background information is contained in the Original FSAR as required for initial plant startup but, per industry practice, is not maintained for the current incumbent personnel in the Updated FSAR. Revised Table 12.1-1 to list all ANSI N18.1-1971 position titles versus PSC position -titles.

Removed the temporary waiver for the Licensed Operator Requalification Training Program as it pertained to instructor qualifications. Revised the Written Procedures section to i reflect current procedures.

Section 13 Stated that the Xenon Stability Test is now allowed to be performed at the maximum achievable steady-state power below 100% 4 reactor power.

Section 14 Revised the loss of normal shutdown cooling description to reflect the EQ Program and shutdown cooling reanalyses which confirmed that a EES cooldown is acceptable for- use during the following scenarios: a) cooling with one water-turbine driven circulator driven by feedwater at normal helium pressure from an

. initial reactor power level of 105%, b) cooling with one water-turbine driven circulator driven by unboosted condensate or boosted firewater from an initial reactor power level of 83.2%,

and c) cooling- with one feedwater driven circulator during helium depressurization at the Maximum Credible Accident (MCA) rate from an initial reactor power level of 105%. Added results of analyses on Safe Shutdown Cooling from 87.5% reactor power with one circulator driven by boosted firewater and one EES supplied with firewater following a 90 minute 10FC. Stated that feedwater cooling via one EES sect 4n will provide adequate cooling following a 30 minute 10FC from 83.2% reactor power.

Included personnel radiological dose rates, equipment radiological doses, and offsite radiological doses due to a worst case HELB. Added that in the event of a steam generator tube or subheader rupture with a wrong loop dump,-utilization of either the intact EES supplied with feedwater or a reheater supplied t _. _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ -

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Attachment A

, to P-87255 with condensate will provide adequate cooling, from an ' initial reactor power level of 100%, after the first 30 minutes. During the first 30 minutes feedwater is postulated to be supplied to the leaking EES. Added equipment radiation doses and personnel radiation dose rates for a MCA, DBA-1, and DBA-2. Stated that cooldown with one circulator driven by feedwater is sufficient for a MCA.. Confirmed that.cooldown for DBA-2 from 105% reactor power with two circulators. driven by feedwater after a 60 minute 10FC can be performed.with feedwater supplying both EES sections.

Stated that the core support floor would perform its safety function during the cooldown following a HELB or DBA-2 without the PCRV liner cooling system operational and the fuel storage wells would safely house irradiated fuel without the aid of any cooling for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Appendix A Added steam generator recrystallization and creep-fatigue final study results. Stated that operating limits have been established for the helium circulators based on helium temperature and circulator speed.

Appendix B For a description of changes to this Appendix see PSC Letter Williams to Calvo, dated July 21, 1876 (P-87261).

Appendix C-Removed the reliance on reheaters for Safe Shutdown Cooling.

Stipulated that the Reactor Building pressure relief system is not an Engineered Safeguard. Stated that SLRDIS initiates a temporary 10FC which is bounded by previous FSAR analysis.

Provided a discussion that SLRDIS sensing and trip logics are different than the normal PPS logic design of 2-out-of-3.

Appendix D Stated the effects on the helium purification system for depressurization during DBA-1. Stated that lower personnel dose rates in the Reactor Building during DBA-1 have been determined based on using the predicted fission product release fractions.

Added a new section on the consequences of a loss of forced circulation from 35% reactor thermal power and 8% reactor thermal power, w __ __ _

, At5achmentA' g ., . to P-87255' Appendix I Added a note stating that the' Steam Pipe Rupture Detection System has been. replaced by SLRDIS ;and gave La brief description' of -

SLRDIS. -Added a.. note . stating that Safe ; Shutdown Cooling, following a 90 minute-10FC. from 87.5% reactor -power, can be

performed using one. firewater pump.to supply one EES section and:

one water'-driven circulator. .Also, stated that cooldown can be performed following a 90 minute 10FC.from 39%. reactor power.using one' firewater pump to supply one reheater > and - one . water-driven

. circulator.. Added a note stating. that: the environmental-qualification requirements of. safe' shutdown.' equipment .is in

-accordance'with 10CFR50.49. Revised Figure I.2-1 to show the new fir _ewater supply line into the emergency condensate header, the e six inch- vents from the EES sections for' Safe Shutdown Cooling, and the check' valve in the emergency-condensate header.

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Appendix J ,

Added. this new ' Appendix for the approved fuel surveillance program.

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. Attachment B~

to P-87255 LIST OF 10CFR50.59 CHANGES NOT-PREVIOUSLY-REPORTED TO-THE COMMISSION As' required, by 10CFR50.71(e)(2)(ii), following is a list

'of the changes made.under-the provisions of 10CFR50.59 which have not been . submitted ;to the Commission. ?These changes will be more. fully described,'along with the.results of their safety evaluations as-appropriate', in.PSC's next annual report of 10CFR50.59 changes.-

FS AR ..

SECTION DESCRIPTION OF-CHANGE 7.1 Revised! Figure 7.1-14l to remove the ~'.'and gate from each channel .of the undervoltage reactor scram inputs 'to' PPS.

8.2 Revised- this section to state that the undervoltage protection system associated with each 480 VAC essential bus utilizes ~an automatic switching feature to- ensure a continuous supply of control. power is available for operability ~ of the undervoltage. relays.. Also. revised this section to add some exceptions to:the electrical wire size requirements. Stated that; overcurrent protection devices are sized for the plant electrical cable installations. Revised Figure 8.2-9 to : reflect installation of a.new power supply cable to achieve at

'least 20 feet cable separation of the Bearing Water Pumps.

Revised Figures-8.2-17 and 8.2-18 to show the new- control power source's .to the main control board (I-10): for selected valves associated with SLRDIS.

9.12 Revised Figures 9.12-1 and.9.12-2 to show the redundant firewater pump discharge cross-connect isolation valves.

10.1 and Revised Figures -10.1-1 and I.2-1 to show the check valve-

I.2 . installed in the emergency condensate header which ensures the integrity.of the ' emergency condensate header when using this path for . Safe Shutdown Cooling following a HELB, the six inch vents from the EES sections for Safe Shutdown Cooling, .and the eight inch vent off of the Bypass Flash Tank.

Various Revised various Sections, Tables , and Figures to incorporate the removal of the Steam Pipe Rupture Detection System and the addition of SLRDIS. This included SLRDIS interface equipment, new components being ,

classified as safety-related, and rerouting of cables as o

!- essential cables to comply with the EQ Program. New Safe

Attachment B

..,' to P-87255 Shutdown Ccoling procedures that ' required FSAR changes have been incorporated into the descriptions related to SLRDIS, the: new six inch vent lines, and the'emergenc feedwater header's removable blind insert (Figure 10.3-9)y .

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gj Attachment C L., to P-87255 RESOLUTION OF NRC COMMENTS CONCERNING REVISION 4 0F THE UPDATED FSAR NRC Letter .Heitner to Williams, dated November 18,1986 (G-86608) identified three discrepancies:

1. -NRC Comment: "Section 3.4.12 - Your response to NRC comment

1. 27 is unclear. Please clarify your internal schedule for justifying replacement of full-length reflector blocks-with half-length blocks.

This should be done in the next FSAR update."

PSC Response: Letter P-86659 (Brey to Berkow dated December 8, 1986) stated that this change wuuld be placed in Revision 5 of the Updated FSAR and that the identified section should be 3 . 4 .1. 2 . Section 3.4.1.2 has been updated to provide the justification for the use of half-length reflector blocks.

2. NRC Comment: "Section 7.3.1.2.1, page 7.3-9, Rev. 4 - This section, as corrected by Rev. 4, is in error since Figure 7.3-20 does not contain a line indicating manual adjustments by steps of the PPS setpoints as indicated in Section 7.3.1.2.1 (refer to Attachment C to P-86466, NRC Comment f

1. 4)."

PSC Response: New Figure 7.3-20 was omitted in error from Revision 4 of the Updated FSAR. The correct Figure 7.3-20 has been placed in Revision 5 of the Updated FSAR.

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' A ..i e Attachment C f m , to P-87255..

. 3. NRC Comment: " Table B.5-4 _ PSC's ' response to NRC Comment #20, L1

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Attachment C to P-86466, indicates that Table B.5-4 is on two sheets'back-to-back. If,this is

a correct statement, please provide us with a-L copy.of this table. . Our.FSV FSAR copy appears to' c be incomplcte."

PSC Response: .As stated in P-86659, both Tables B.5-3 and B.5-4 were comprised of two sheets back-to-back and -13 L,.

copies of Table B.5-4 were attached to P-86659.

' Tables B'.5-3 and B.5-4; were again updated by Revision 5 of the Updated FSAR.

, A telephone conversation from the NRC's Heitner, Stacbeu, land i Shemanski to PSC's ' Holmes , et. - al . (C-86-0114). identified the- 'l following two. discrepancies:

4. N_RC Comment: Section.7.1.2.6. states. that tripping a circulator results in reducing the load setting of the main turbine. governor -to 65% load and P-85214 states'  !

that the-load is reduced to 50%.  !

j PSC Response: Section .7.1.2.6, has been updated to reflect the correct value of 50%.

5.- NRC Comment: - Table 14.5-3 values listed for. " Total H20 Inleakage" for cases 2, 4, and 6 don't correspond- l to the steam content values in Figures 14.5-2,  !

14.5-4, and.14.5-6. j

-PSC Response: Table 14.5-3 has been updated to reflect the. i correct values for " Total H20 Inleakage." This j table.was also updated to reflect changes made to Section 14.5 and figures in Section 14.5 based on >

the verification of non-Safe Shutdown Cooling using a EES or reheater.

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