Text

Forwards Response to Redrafted Tech Specs for Helium Calculators,Steam Generators & Pcrv Liner Cooling Sys.Review of Encl Issue 2 to CMG-4 & Meeting Requested to Resolve Any Remaining Comments on Specs
	ML20137X211
Person / Time
Site:	Fort Saint Vrain
Issue date:	02/28/1986
From:	Brey H; PUBLIC SERVICE CO. OF COLORADO
To:	Berkow H; Office of Nuclear Reactor Regulation
Shared Package
ML20137X216	List: ML20137X211; ML20137X219;
References
	P-86169, NUDOCS 8603050288
	Download: ML20137X211 (46)
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OPublic Service ~

4 ogwnpery orcenso 2420 W. 26th Aver,te., Suite 1000, Denver, Co:orado 80211 Februarj 28,198E Fort St. Vrair Unit No. 1 P-86]69 Direct'.:r tf Nucleae Reactor llegulation U.S. Nuclear P.egulatcry Cormlit;sion Washing'.tn, D, 6 20535 l

ATTN: Mr. H. N. Berkow, Director Standardi ation and 3pecio Prnjects Direr;torete Dociet No. 50-267 l

SUBJE6f: Technical Specification Upprade Fecgram Redrafted LC0's REFEREtJCES: 1)NRCletterdated 12/27/85, Berkow to Valktr (G-06008)

2) PSC lecter dated 2/7/C6, Brey to Berkow(P-86095)

3) PSC letter dated f

10/11/85, Brey to Cutcher (P-A5363)

Dear Mr. B&rlow:

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Attached is PSC's response to the re. drafted Technical Specificaticals for the helium circulators, steam generators, and PCRV liner cooling system et Fort St. Vrain (FSV), as provided by the hRC f;i Reference 4

1.

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As ir,df cated in Reference 2 and as discussed with Mr. G, t.. Plumlee 10 (NRC), PSC has rewritten tne Reference 1 Specifications to

r. ore correttly reflect the FSAR counitnents, arti to apply Standard Technical Specification auidance in a manner that is appropriate for comparable tWiprent.

Att6chr.ent 1 is a distcsrion of these issues trd of other concepts that PSC has used in the FSY Technical Specification Ilograce Progrant, reummm P

Ppit

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'. contains the subject Specificatior.s as rewritten by PSC. is a re-submittal of PSC's response to the applicable action items. Attachment 4 is a discussion in support of the use of a calculated bulk core temperature of;760 degrees F to indicate when

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redundancy in safe shutdown cooling equipment is required; this discussion also responds to NRC questions identified in a telephone conversation on February 6, 1986.

PSC requests that the NRC review Attachment 2 and schedule a raeeting to resolve any remaining comments on these Specifications.

1 If you have any questions regarding the attached information, please

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contact Mr. M. H. Holmes at (303) 480-6960.

Very truly yours, 4

vC 1

I H. L. Brey, Manager J

hu;1 ear Licensing and Fuels l

HLB /SWC/ paw 1

Attachments cc: Mr. G. L. Plumlee III (NRC) i t

a 1

b To P-86169 DISCUSSION OF CONCEPTS USED IN FSV TECHNICAL SPECIFICATION UPGRADE PROGRAM A.

General Throughout the Technical Specification Upgrade Program (TSUP),

PSC has revised the existing Specifications to make them clear, concise, correct and consistent with the licensing basis of Fort St. Vrain, as embodied in the FSAR. The guidance contained in ANSI /ANS 58.4, and as exemplified in the Standard Technical Specifications, has provided the basis for determining what requirements are included in the FSV Specifications.

FSV systems have been compared to LWR systems to determine overall Tech Spec organization, level of detail, and appropriate action times.

For example, the Safe Shutdown Cooling Systems at FSV are analogous to the ECCS systems at an LWR, and the same requirements for redundancy and action times have been applied at FSV.

Also, in accordance with the initial work specification provided to the NRC (WS-TS-1, Attachment 1 to PSC letter P-84530, dated 12-14-84, Lee to Johnson), any significant analyses or plant modifications are outside the scope of the TSUP, and would be treated as separate licensing issues. Requirements S r equipment outages due to surveillances and maintenance are therefore a consideration in establishing action times, to the extent that they are consistent with plant safety bsed on the small likelihood of system failures during the allowable outage times.

B.

OPERATING Equipment With these concepts in mind, PSC considers that the Technical Specifications should contain requirements for system conditions to the extent that they are relied upon in the safety analysis (per ANSI /ANS 58.4).

Therefore, a requirement for OPERATING primary coolant

loops, with numbers of OPERATING helium circulators corresponding to various reactor power levels, is not essential in the Tcchnical Specifications, as it does not reflect any FSAR requirement.

The upgraded Technical Specifications ensure forced circulation through the Power-to-Flow Specification (3/4.2.6) and through the Minimum Helium Flow Specification (3/4.2.4) for power levels below 25% rated thermal power.

Secondary system flow is ensured by the Reheat Steam Temperature High scram function of the Plant Protective System Specification (3/A.3.1).

C.

Single Loop Operation Single loop operation is a design feature of FSV, as discussed in FSAR Section 4.3.5.2.

However, FSAR Section 4.3.4 indicates that with one primary coolant loop inoperable, operation on the one remaining loop may continue long enough to perform an orderly shutdown.

The actions for inoperable primary coolant loops have

.been revised to reflect this orderly shutdown requirement.

D.

Redundancy for DBA-2 With the singular exception of the Rapid Depressurization Accident (DBA-2), one circulator on pelton wheel drive assures safe shutdown cooling, and requiring two circulators to be operable provides protection against a single failure.

In the depressurized condition after DBA-2, the less dense helium requires that twp circulators be operating on 8000 rpm pelton wheel drive to assure safe shutdown cooling.

For this case, requiring two operable circulators does not allow for a single failure.

This is acceptable because of the extremel'y low probability of the occurance of DBA-2. A probabilistic analysis performed by GA Technologies has indicated that the probability of a DBA-2 depressurization is less than 1 x 10-9 per year with an uncertainty factor of much less than 90.

This analysis is included at the end of this discussion. This is much less than the safety goal currently being considered by the NRC.

The proposed Technical Specifications assure sufficient operable ecuipment to provide safe-shutdown cooling for this low probability event.

PSC does not consider that it is appropriate to require additional operable equipment to allow for an additional failure.

The FSAR (Section 10.3.10, Safe Shutdown Cooling with Single Failures in Cooling Water Supplies), does not indicate that single failure protection is provided for DBA-2, where two circulators are required. PSC considers that this degree of protection is not a feature of the FSV design or licensing basis.

E.

Circulator Performance Requirements The capability of a helium circulator to provide 4.5% flow with condensate drive for the pelton wheels is a design feature of FSV that is not relied upon in the safety analysis.

Since firewater is the only Class I safety-related pelton wheel drive source relied upon for accident analysis, PSC does not consider that a 4.5% flow demonstration is appropriate in the Tech Specs.

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F.

Liner Cooling System - Firewater Connection 3

Similarly, the capability of providing firewater to the PCRV liner cooling system is a design feature of FSV that is not relied upon in the safety analysis.

Liner cooling system redundancies assure system operability and PSC does not consider that the firewater connection should be included in the Tech

. Specs.

G.

Circulator Auxiliary Ecuipment Equipment such as the emergency water booster pump and. turbine water removal pump are used to ensure safe shutdown cooling drive capability for the helium circulators and are considered as safe shutdown cooling equipment. As such, the inoperability of one of thesd items (e.g., ' emergency water booster pump) constitutes a loss of redundancy in a safe shutdown cooling system.

It does not render a circulator inoperable, and a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action 4

statement is consistent with an STS ECCS inoperability.

This 3

feature 13 included in Attachment 2.

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Attachment to GP-2700 1

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From:

L. L. Parme IN REPLY REFER TO: SR:LLP:105:35 To:

A. J. Kennedy DATE: 8 November 1985 3UBJECT: Frequency of Large Leaks in FSV Reference 1: " Fort St. Vrain Nuclear Generating Station Final Safety Analysis Report," Public Service Company of Colorado Reference 2: 'An Assessment of the Integrity of ?WR Pressure-7essels,'

Report by a Study Group under the chai.manship of Dr. W.

Marshall C3E, FRS. Oct 76 Reference 3: "RTGR Accident Initiation and Progression Analysis Status Report - VolIII," GA-A13617, Nov 1975 Summarv For the Fort St Train Nuclear Generating Station safety analysis (Ref.

1) a rapid primary coolant depressurization is included.

This hypothetical accident, designated DBA 2, consists of the simultaneous failure of two closures in a PCR7 penetrat. ion resulting in a rapid i

primary coolant loss.

The likelihood of this accident has been e s timated to be approximately 1 x 10-9 per year with an uncertainty factor (P95/ Median) of much less than 90.

Dicussion or the various penetrations in the FS7 PCR7, 57 are capable of leading to substantial depressuri:ation rates if they were to fail.

These are the 12 steam generator module, 37 refueling, 1 top head access, 1

bottom head access and 4 circulator penetrations.

The evaluation in the FS7 FSAR for DBA 2 consists of postulating and analyzing the callure of the closures for thesa ~ penetrations.

A commonly quoted median failure rate of a nuclear pressure vessel is 10-7 per vessel-year with an uncertainty factor of about 10 (Ref. 2).

Since such a vessel may contain on the order of 100 penetrations, the probability of a single structual failure in any given penetration cannot be signi.fic an tly greater th,an 10-9 per year.

This failure frequency is used as an estimate for th e likelihood of failure for any single primary closure on th e FS7 PCRV, all of which are designed in accordance with the principles of the ASME Boiler and Pressure 7essel Code,Section III., Class A.

The failure of a single closure, however, is not enough to cause a transient such as that described by DBA 2.

Each penetration of the PCRV is closed by two independently secured closure structures.

Thus a failure of these two closures must occur before such a leak can be experienced.

Significant effort has been expended to assure independence in the design of these closures.

Continuous leak monitoring and vertical movement stops are two examples.

Nevertheless, they are subject to a variety of common factors such as I

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_2 in a t e r ials, welding, environment, etc. which could possibly lead to

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common mode failure.

While quantification of this commen mode factor could be somewhat difficult a conservative value, representative of active components, can be used as an estimate.

In this seso a 3 eta factor of 0.02 (rer 3) is assu=ed between th e redundant penetration closures.

Th e re f o re,

taking the 57 penetrations included within D31 2,

a penetration failure frequency of 10-9 per year and the probability that th e second closure fails given th a t a first fails, the median frequency of a DBA 2 type depressurination occurring is about 1x 10-9 per year.

Prob = (57 penetrations) x 10-9/7 ear x 0.02

= 1 x 10-9 per year In a 30 year plant life this corresponds to a 3x10-8 probability of such an accident occuring.

The uncertainty of this es timate is due to both the uncertainty in the assessed frequency of a single closure failure, 10, and that of the co==on mode failure fac tor between closures.

While the assumed use of a Beta factor characteristic of active, powered components is very likely conservative, there is significant uncertainty.

However the common mode factor cannot be greater than unity bounding the common mode uncertainty at 50.

So, assuming legnormal distributions, the total uncertainty for this estimate is less than 90.

cc:

C. J.

Everline F.

A.

Silady I have read th e s e failure frequency analysis and results and J

find it acceptable.

AP C. 'J.

Everline

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e ATTACHMENT 2 TO P-86169 i

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Amendment No.

Page 3/4.5-DRAFT SAFE SHUTDOWN COOLING SYSTEMS

. FEB 2 'c mr5 3/4.5.1 SAFE SHUTDOWN COOLING EQUIPMENT LIMITING CONDITION FOR OPERATION 3.5.1.1 a.

Two primary coolant loops shall be OPERABLE, each with at least:

1.

One helium circulator OPERABLE, and 2.

One steam generator section (either the economizer-evaporizer-superheater (EES) or the reheater)

OPERABLE.

b.

For OPERABLE helium circulators, the following safe shutdown cooling drives and auxiliary equipment shall be OPERABLE:

1)

A safe shutdown cooling drive with the capability of providing the equivalent of 8000 rpm circulator speed at atmospheric pressure to two circulators simultaneously, c-5," f. r 2)

Two safe shutdown cooling drives with the capability s

grWU of providing 3% rated helium flow at operating Q,'M

.J pressure with firewater

supply, including two hg),(,,

OPERABLE emergency water booster pumps and two i

OPERABLE flow paths, 3)

The turbine water removal system shall be OPERABLE, including two turbine water removal pumps, 4)

The normal bearing water system, including two sources of bearing water makeup and two bearing water makeup pumps (P2105 and P2108),

5)

The associated bearing water accumulator (T-2112, T-2113, T-2114, or T-2115), and

Amendment N3.

Page 3/4.5--

bSAFT FEBpv;qy 6)

The supply and discharge valve interlocks ensuring automatic water turbine start capability.#

APPLICABILITY:

POWER, LOW POWER, STARTUP* and SHUTDOWN

ACTION:

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With only one OPERABLE primary coolant loop, and with both a.

loops OPERATING, restore the inoperable loop to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

With only one OPERABLE and OPERATING primary coolant loop, restore the non-operating loop to OPERATING status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , or be in at least SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Restore two loops to OPERABLE status within-72 hours of the initial loss of an OPERABLE loop or be in at least SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , c.

With no OPERABLE safe shutdown cooling circulator drive capable of 8000 rpm equivalent, restore the required drive to OPERABLE status within the next 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or be in at least SHUT 00WN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d.

With only one OPERABLE firewater-supplied helium circulator drive, restore at least two drives to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

e.

With only one OPERABLE turbine water removal pump, normal bearing water system, source of bearing water makeup, or bearing water makeup

pump, restore the inoperable equipment to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or be in at least SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

f.

With less than the above required OPERABLE bearing water accumulators, restore the inoperable equipment to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or be in at least SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Whenever CALCULATED BULK CORE TEMPERATURE is greater than 760 degrees F.

The supply and discharge valve interlocks ensuring automatic water turbine start capability are only required to be OPERABLE in POWER.

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Amendment N3.

Page 3/4.5-DRApr FEB 2 S ses g.

With less than the above required OPERABLE valve interlocks, restore the inoperable equipment to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or be in at least LOW POWER within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

h.

With loss of both redundant OPERABLE components or flow paths required by LC0 3.5.1.1.b.2, 3, or 4,

(e.g.,

with both turbine water removal pumps or with both emergency water booster pumps inoperable),

be in SHUTDOWN immediately.

1.

With no OPERABLE or OPERATING primary coolant loops, be in at least SHUTDOWN immediately and restore at least one loop to OPERATING status within 90 minutes, or depressurize the PCRV in accordance with the applicable requirement below.

If forced circulation is restored within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months of initial loss, depressurization may be discontinued.

1.

As a function of reactor THERMAL POWER equal to or greater than."5% prior to SHUTDOWN, as delineated in Figure 3.4.4-1.

2.

As a function of CORE AVERAGE OUTLET TEMPERATURE for reactor thermal power less than 25%

prior to SHUT 00WN, as delineated in Figure 3.4.4-2.

3.

As a function of time from reactor SHUTDOWN as delineated in Figure 3.4.4-3.

SURVEILLANCE REQUIREMENTS 4.5.1.1 a.

The helium circulators shall be ' demonstrated OPERABLE:

1.

At least once per 31 days by testing the bearing water accumulators and verifying accumulator flow to the circulator bearing.

2.

At least once per 92 days by:

a)

Performing a turbine water removal pump start test based on a simulated drain tank level to verify automatic actuation and pump start capability, and

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Amendment Ns.

Page 3/4.5-bi3/i T

'O FEB2 s ngg b)

Performing a start test of bearing water makeup pump P-2105, based on a simulated low pressure in the backup bearing water supply line to verify automatic actuation and pump start capability.

3.

At least once per REFUELING CYCLE:

a)

Testing the water turbine inlet and outlet valve interlocks ensuring automatic water turbine start capability by simulating a steam turbine trip, and b)

Monitoring the proper closure of the circulator helium shutoff valves.

c)

Performing a functional test of each emergency water booster pump.

4.

At least once per REFUELING CYCLE on a STAGGERED TEST BASIS whereby circulators 18 and ID will be tested during even numbered cycles and circulators 1A and 1C during odd numbered

cycles, by demonstrating operation on water turbine drive by:

a)

Verifying an equivalent 8000 rpm (at atmospheric pressure) on two circulators si.aultaneously on feedwater motive power using t.he emergency feedwater header, and b)

Testing each circulator by verifying equivalent 3% rated helium flow on condensate at reduced pressure (to simulate firewater pump discharge) using each emergency water booster pump (P-2109 and P-2110).

5.

At least once per 10 years by verifying:

a)

A helium circulator compressor wheel

rotor, turbine wheel and pelton wheel are free of both surface and subsurface defects in accordance with the appropriate methods, procedures, and associated acceptance criteria specified for Class-I components in Article NB-2500,Section III, ASME Code. Testing shall be scheduled so that over 4 inspection periods, each circulator will be tested once. Other helium circulator components, accessible without further disassembly than-required to inspect these wheels, shall be visually examined, and

Amendment No.

Page 3/4.5-DRe m FEB e b)

At least 10% of primary coolant pressure boundary bolting and other structural bolting which has been removed for the inspection above and which is exposed to the primary coolant shall be nondestructively ~ tested for identification of inherent or developed defects.

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c)

Reports 4

Within 90 days of examination completion, a Special Report shall be submitted to the NRC in accordance with Specification 6.9.2.

This report shall include the results of the helium circulator examinations.

b.

The steam generators shall be demonstrated OPERABLE:

1.

At least once per 18 months be verifying proper flow through the emergency feedwater header and emergency-condensate header to the steam generator sections.

2.

At least once per 5 years by volumetrically examining.

the accessible portions of the following bimetallic welds for indications of subsurface defects:

a)

The main steam ring header collector to collector drain-piping weld for one steam generator module in each loop, and b)

The same two' steam generator modules shall be re-examined at each interval.

The initial examination shall be performed during SHUTDOWN or REFUELING prior to the beginning of fuel cycle 5.

This initial examination shall also include the bimetallic welds described above

.for two additional steam generator modules in each loop.

3.

Tube Leak Examination Each time a steam generator tube in plugged due to a leak, specimens from the accessible subheader tubes connected to the leaking inaccessible tubes shall be metallographically examined.

The result, of this metallographic examination shall be compared to the results from the specimens of all previous tube leaks.

Amendment No.

Pago 3/4.5-DRAFT FE8P P ex A study shall be performed to evaluate the size and elevation of the tube leaks to determine if a cause of the leak or a trend in the, degradation can be identified.

a)

Acceptance Criteria

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An engineering evaluation shall be performed to determine the acceptability of:

1)

Any subsurface defects identified in Specification 4.5.1.1.b.2, 2)

Continued operation considering the condition of the steam generator materials, and 3)

OPERABILITY of the steam generator sections considering the number of plugged tubes and their ability to remove decay heat.

b)

Reports Within 90 days of the return to operation following each steam generator tube leak study a Special Report shall be submitted to the Commission in accordance with Specification 6.9.2.

This report shall include the estimated size and elevation of the leak (s), and at least the preliminary results of the metallographic and engineering analyses performed, the postulated cause of the leak if identified and corrective action to be taken.

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l TEN AVAILASLE PR80R TO MfflAT198 0F PCRV SEPRENU442AT195 WNE5 FORCEO CIRCULATION E M FW A MMO NWR AT M a

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20 30 40 50 80 70 IO 90 100 REACTOR THERMAL POWER -%

Tame Avainbie Prior to Initiation of PCRV Depmemustion When Forced Circulation is Last from a Powered Condition at FSV l

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A 400 500 800 700 000 900 1000 1100 1200 1300 1400 1500 AVERAGE CORE OUTLET TEllPERATURE 0F Time Available Prior to Initiation of PRCV Depressurization as a Function of Aversee Core Outlet Temperature at the Onnet of a LOFC Figure 3.5.1-2 l

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CIRCULATION 88 LOST FR0tl A SHUT 00WN 9

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TO BE USEO FOR A SHUT 00WN CONDITION ONLY 8

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0 100 200 300 400 500 600 700 800 900 1000 TIME FROM REACTOR SHUT 00WN - HOURS Tune Available Prior to Initiation of PCRV Depressurization When Forced Circulation is Lost from a Shut Down Condition Figc e 3.5.1-3

Amendment N3.

Paga 3/4.5-SAFE SHUTDOWN COOLING SYSTEMS 3/4.5.1 SAFE SHUTOOWN COOLING EQUIPMENT g fz.,

r FE8 2 LIMITING CONDITION FOR OPERATION 3.5.1.2 a. At least one primary coolant loop shall be OPERABLE, including at least:

1.

One helium circulator OPEPABLE, and 2.

One steam generator section (either the economizer-evaporator-superheater (EES) or reheater) OPERABLE.

b. For at least one OPERABLE helium circulator, the following emergency drives and auxiliary equipment shall be OPERABLE:

1.

One safe shutdown cooling drive with the capability of providing the equivalent of 8000 rpm circulator speed at atmospheric pressure, 2.

One safe shutdown cooling drive with the capability of providing 3% rated helium flow at operating pressure with firewater supply, including one OPERABLE emergency water booster pump and one OPERABLE flow path, 3.

The turbine water removal system, including one turbine water removal pump, FOR 4.

The normai eeerieg water system, inciuding eee soerce or bearing water makeup and one bearing water makeup pump h(n.'

(P2105 g P2108), and 5-

'**"'"S

"*" " " "*" " ('-2112, 1-2113. T-2114 C -

E T-2115) for the OPERABLE circulator (s).

APPLICABILITY:

STARTUP*, SHUTDOWN *, and REFUELING Whenever CALCULATED BULK CORE TEMPERATURE is less than or equal to 760 degrees F.

Amendment No.

Page 3/4.5-DRAF:

FEB 2 819" ACTION: a. With less than the above required OPERABLE equipment, and with forced circulation maintained, be in at least SHUTDOWN immediately, and restore the required equipment to OPERABLE status prior to reaching a CALCULATED BULK : C.0RE TEMPERATURE of 760 degrees

.F, or suspend all operations involving CORE ALTERATIONS or control rod movements resulting in positive reactivity changes, or movement of IRRADIATED FUEL.

b. With less than the above required OPERABLE equipment, and with no OPERATING primary coolant loops, be in at least SHUTDOWN immediately and restore at least one loop to OPERATING status prior to reaching a CALCULATED BULK CORE TEMPERATURE of 760 degrees F, or 1.

Suspend all -operations involving CORE ALTERATIONS or control rod movements _ resulting in positive reactive changes, or movement of IRRADIATED FUEL, and 2.

Ini'.iate PCRV depressurization in accordance with the time specified in Figures 3.5.1-2 or 3.5.1-3, as applicable.

SURVEILLANCE REQUIREMENTS 1

4.5.1.2 No additional Surveillance Requirements beyond those specified in SR 4.5.1.1.

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Amendment No.

Page 3/4.5-DRU" FEB 2 r -

SASIS FOR SPECIFICATIONS LCO 3.5.1.1/SR 4.5.1.1 AND LC0 3.5.1.2/SR 4.5.1.2

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One primary coolant loop, consisting of one helium tirculator and one steam generator section ensures SAFE SHUTDOWN COOLING when the plant is pressurized. Two loops are specified during POWER, LOW POWER, STARTUP, and SHUTDOWN with CALCULATED BULK CORE TEMPERATURE greater than 760 degrees F to allow for a single failure in either the heat removal equipment or circulator auxiliary equipment which provides services to one loop. One circulator, operating with motive power from either (a) condensate or boosted firewater supplied via the emergency condensate header, or (b) feedwater or boosted firewater supplied via the emergency feedwater header, provides sufficient primary coolant circulation for the pressurized condition. SAFE SHUTDOWN C0OLING is discussed in FSAR Section 10.3.9, single failure considerations in Section 10.3.10, and condensate and boosted firewater cooldown transients in FSAR Sections 14.4.2.1 and 14.4.2.2.

It is important to note that a distinction is made between OPERABLE and OPERATING equipment.

A circulator may be OPERATING on normal steam drive, but not be OPERABLE from the standpoint of Safe Shutdown Cooling, by virtue of a failed pelton wheel inlet valve.

Similarly, a circulator may be OPERABLE from the standpoint of being capable of operating on pelton wheel drive, but it may not be OPERATING.

Two circulators, operating with emergency water drive, supplied with feedwater via the emergency feedwater header, provide sufficient primary coolant circulation following a postulated Design Basis Depressurization Accident (DBA-2).

DBA is a highly incredible event and protection against single failures is not a feature of FSV.

The requirements for OPERABLE steam generator (s) provide an adequate means for removing heat from the primary reactor coolant system to the secondary reactor coolant systen. The helium flow which cools the reactor core enters the steam generator at high temperature and gives up its heat to the reheat steam section and main steam / water section.

Each steam generator consists of six identical individual steam generator modules operating in parallel.

Each module consists of a reheater section ano an economizer-evaporator-superheater section. Any one section provides sufficient heat removal capability to ensure SAFE SHUTDOWN COOLING.

Amendment No.

Page 3/4.5--

l DRAFT e

FEB 2 8198S During POWER, LOW POWER, STARTUP and SHUTDOWN with CALCULATED BULK CORE TEMPERATURE greater than 760 degrees F,

one steam generator' section in each loop is required to be OPERABLE.

This allows for a single failure and provides art adequate means for removing decay heat.

During SHUTDOWN with CALCULATED BULK CORE TEMPERATURE less than or equal to 760 degrees F and REFUELING,- redundancy is not required-and either the reheater section or the economizer-evaporator-superheater section of one steam generator can be used for shutdown heat removal from the primary coolant.

Safe Shutdown Cooling Drives j

The requirement to be able to drive two circulators

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simultaneously at 8000 rpm,. at conditions equivalent to j ~

atmospheric pressure, ensures Safe Shutdown Cooling capability in the event of DBA-2. This drive capability uses feedwater, supplied via the emergency feedwater header.

Testing to demonstrate capability at conditions equivalent to atmospheric pressure is based on calculated helium density, reactor-pressure and circulator inlet temperature. A 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action statement rectoration time is acceptable because DBA-2 is a highly incredible event, with a probability of occurance of less than E-7 per year.

l The requirement for a drive capable of 3% rated helium flow with firewater supply ensures Safe Shutdown Cooling in the event of a loss of normal circulator steam turbine drive,.in j

the pressurized condition.

A 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action statement

. restoration time is used consistently in this-Specification.

j for loss of redundancy in the Safe Shutdown Cooling Systems.

4 This is acceptable in that an equipment failure in this j

relatively brief interval is unlikely.

In the event of a complete loss of Safe Shutdown Cooling capability, an i

immediate shutdown is required, to place the -reactor _ in a stable condition with reduced decay heat. loads.

t Circulator Auxiliary Equipment i

One turbine water removal pump has sufficient capacity to remove the water from two circulator water turbines.

Also, i

i the turbine water removal tank overflow to the reactor i

building sump is available if the normal pump flow path is lost.

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Amendment No.

Paga 3/4.5-DRAFT FEB 2 8198S Each independent-bearing water system provides a continuous supply of bearing water to the two ettculators in each primary cooling loop.

A backup supply of bearing water is provided from tne steam generator feedwater system.

Makeup. bearing water requirements are also nor:nally obtained from the feedwater system. A separate bearing water makeup pump is provided as a backup to supply makeup water to the bearing water surge tank. The bearing water makeup pucp normally takes suction from the deaerator but can also be supplied from the condensate storage tanks.

If this pump is fooperative, an emergency bearing water makeup pump can supply water at a reduced capacity from the condensate storage tank to the bearing water surge tank.

In an extreme emergency, filtered firewater can be provided to the bearing water surge tenk by either the baaring water makeup pump or'the emergency bearing water makeup purry.

Each bearing water locp contains a gas pressurizer and bearing water accumulator capable of supplying bearing water for 30 seconds at design flow rate if no other source of bearing water is available, This is adequate for shutdown of the affected circulators without damage to the bearings.

The bearing water system, includir.g the bearing Water accumulators and the bearing water makeup pumps are functionally tested at 31 days and REFUELING CYCL,E intervals, respectively, to ensure proper operation.

There is no redundancy in the bearing water accumulators and a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months action statement restoration time is provided.

This is acceptable considering the low likelihood of failures of these ccmponents.

Auto water turbine $ tart.is prevented if a water turbine TRIP exists or the auto water turbine start control switch is not in the auto position. The aforecentioned interlock circuitry is tested once per REFUELING CYCLE, to ensure proper system operation.

The automatic water turbine start feature is relied upon in the event the control room has to be abandoned.

Since this is an unlikely event, a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action statement restoration time is acceptable.

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Amendment No.

Page 3/4.5-DRAFT Depressurizatjon FEB 2 6 m In the unlikely event that all forced circulation is lost for 90 minutes, start of depressurization is initiated as a furiction of prior power levels, with 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months from 100% RATED TfiERMAL POWER being the most limiting case.

Operators will continue attempts to restore forced circulation cooling uFtil 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after the loss of forced circulation.

Multiple sources and flowpaths to eistablish forced convection cooling using circulators makes required depressurization highly unlikely.

Cooldown using forced circulation cooldown is preferred to a depressurized cooldown with the PCRV liner cooling system.

Depressurization of the PCRV under extended loss of forced circulation conditions is accomplished by venting the reactor helium through a train of the helium purification system and the reactor building vent stack filters to atmosphere. Start of depressurization tirnes from various reactor power conditions are delineated in Figures 3.5.1-1, 3.5.1-2, and 3.5.1-3 and are discussed in the FSAR Section 9.4.3.3 and Appendix D, Redundancy Criteria The use of 760 degrees F CALCULATED BULK CORE TEMPERATURE as a division between the APPLICABILITY of Specification 3.5.1.1 verses 3.5.1.2 is explained as follows:

In the FSV HT6R, the limiting parameter of interest is a core inlet tenperature greater than 760 degrees.F. The CALCULATED BULK CORE TEMPERATURE is a conservative calculation of the maximum potential temperature fr the coro and surrounding components. The conservatisms tre such that if the CALCULATED BULY, CORE TEMPERATURE is limited to 760 degrees F, the design inlet temperature of 760 degrees F is not exceeded. Systems used for accident prevention and mitigation are reoired to satisfy the single failure criterion whenever CALCULATED BULK 00RE TEMPERATURE is greater than 750 degrees F.

However, When CALCULATED BULK CORE TEMPERATuliE is equal to or less than 760 degrees F, it is acceptable to require only one OPERABLE system for accident prevention and mitigation without single failure consideration, on the basis of the limited core cooling requirements.

Am;ndment No.

Pack 3/4.5-DRAFT FE0 2 81996 All forced circulation may be interruoted for mafntenance purposes previoed that the time calculated fer CALCULATED GULK CORE TEMPEPATURE to reach 760 #grees F is not :exsdeded.

However, if forced circulation b teaporarily restored, a recalculation shall be performed, toued on Dresent conditions, to establish a new time period for CALCULATED BULK-CDP.E TEMPERATURE to redch 760 degrees F.

Redundant systems thay alto be taken out of service for inaintenance or serveillance testing provided that fcredd circulation is maint6ined. The time to ree;;h CAL.CULATED B' LK CORE TEMPERATURE equal to 750 J

degrees F may be recalculated as t)ften as required.

Steam Generators The steam generatcr reheater er EES secti6r.3 can receive wat.er from either the essociated energency condensate heater or trie emergency feedwatcr header which are requireri to be OPERABLE per this Specification. System flow C.cEkACILITY is detennined by verifying ficw frcm each of the aforementioned EmePgency headers through each steam generator.

Bimetallic Weld Examinatien The stean generator crossover tube bimetallic welds between Inculoy 800 od 21/4 Cr-1 Mo materiats are not accessible far examination.

The bimetallic welds between steam generator ring header collector, the main steam piping, and the collector drain piping are accessible, involve the Mme materials, and operate at conditions not significantly different from tne crossover tube hixetallic welds, The collector drain pipf g weld is also geonatricdly similar to the crossover tube Wald.

Although minimal degradation is expected to occur, this specification allows for detection of defects which might result from conditions that can unic;ue'ty affect bimetallic relds made between these materials.

Additional collector welds are inspected at the initial examination to establish a baseline which could be used, should defects be found in later inspections and additional examinations subsequently be required.

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Anandment No.

Page 3/4.5-DRAFT Tube Leak Examination gg 2 g pgej Ouring the lifetime Of the plant, a certain number of htsam generator tube leaks are expected to occur, tad the steam generators have been desigfled to have these leaking tube subheaders plugged without affecting the plant 's performance as shown in FSAR Table 4.2-5.

The consequences of steam generator tube leaks have been analyzed in FSAR Section 14.5.

It is ipp6rtant to ipeittify the approximate size and elevation of steam getterator tube leaks and to metallographically examine the subheader tube material because this information 4

can be used to analyre any trend or generic cause of tube

leaks, Conclusive identification of the cause of a steam generator tabt leak niay enable modifications and/or changes in operation to increase the relichility and M fe of the steam generators and to prevent -a quantity of tube failures in exces; of tnose analyzed in the FSAR.

Because of the sutheader designs leading to the steam gener6 tor tube tendles, interna) or external inspection and eys10ation of a tube leak to establish a conclusive cause is not practical, Metallographic examination of the accessible connectibg subheadce tube will show the condition of the internal subheader wall, giving an indication of the ctnditions cf the leaking tube internal well.

thereby der;nnstrating the EffectiV6nGss of Water Chemistry controls.

Determining the ecproximatG size and elevation of the tube leak.may enable evaluation of other possible leak causes such as tube / tube support plate inttirface effects.

The serve 11'ance plan outlined above is considered Edequate to evaluate steam generator tube integrity and ensure that the conset;uences of postulated tube leaks remain witnin tha limits analyzed in the FSAR.

4

Amendment N).

Pag 2 3/4 6-t DRAFT PCRV AND CONFINEMENT SYSTEMS FEB 2 81935 3/4.6.2 REACTOR PLANT COOLING WATER /PCRV LINER COOLING SYSTEM -

OPERAT_ING I.IMITING CONDITION FOR OPERATION 1

m 3.6.2.I The Reactor Plant Cooling Water (RPCW)/PCRV Liner Cooling Systen (LCS) shall be OPERABLE with:

a.

Two (2) loops OPERATING each with at least one heat exchanger and one pump OPERATING; b.

At least three (3) out of any four_(4) adjacent tubes on the core support floor side Wall, core support floor bottom casing, PCRV cavity liner sidewalls and PCRV Q

cavity liner bottom head shall be OPERATING;

)t c.

At least five (5) out of any six (6) adjacent tubes on 1is the PCRV cavity liner top head and core support floor top casing shall be OPERATING, and d.

Tubes adjacent to a non-operating tube shall be l

OPERATING APPLICABILITY-POWER, LOW POWER, STARTUP* and SHU70OWN*

ACTION a.

With only. one (1) RPCW/PCRV LCS loop OPERATING, ensure both heat exchangers are OPERATING in the OCERATING loop, restore the second loop to OPERATING within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in SHUTDOWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and suspend all operations involvieg positive reactivity 1

changes. Without both heat exchangers in the OPERATING loop OPERATING or withow',any liner Cooling system loop flow te in SHUTDOVN within 15 minutes and. suspend all operations involving control red movenients resulting in positive reactivity changes.

Wherever CALCULATED BULK CORE TEMPEP.ATURE 15 greater than l

7o0 degrees F.

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Amendment Ns.

-Page_3/4 6-DR!v FEB 2 8 Se6 b.

With less than the above required number of PCPV Liner Cooling System tubes OPERATING, other than as in ACTION

a. above, restore the required tubes to OPERATING status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in SHUT 00VN within the fel. lowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and suspend all operations inyelving positive reactivity changes.

SURVEILLANCE REQUIREMENTS 4.6.2.1 The RPCW/PCRV Liner Coo 11 rig System shall be demonstrated OPERABLE:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , by verifying that each PCRV Liner Cooling System loop is circulating cooling water

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at a flow rate greater than 1100 gym.

b.

At least once per 31 days, by verifying that liner cooling tube outlet temperature readings and their respective inlet header temperatures (for an operating loop) are within one of the following limits:

1.

30 degrees F temperature rise for tubes cooling top head penetrations; 2.

20 degrees F temperature rise for all other zones except tubes specified below; 3.

Exceptions a) Core Outlet Thermometer Penetrations Tube Delta T 7593 23 degrees F b) Core Barrel Seal / Core Support Floor Area Tube Delta _T F12T46 47 degrees F F7T43 39 degrees F F6T44 43 degrees F F11T45 38 degrees F F5T47 46 degrees F 4

Amendment Ns.

P.:ge.3/4 6-

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DhhP'f c)-Peripheral Seal IES28lgg3 Tube Delta T i

J 3S9 23 degrees F' 45188 23 degrees F 4S10 23 degrees F 3S187 23 degrees F If the tube temperature rise for any liner cooling tube.

is not available due to an instrument failure, the_ tube may be considered OPERABLE if two tubes on both sides of the tube with att instrument failure (4 tubes total) are within their respective temperature lir. tits as _ specified above.

c.

At least once per REFUELING CYCLE by:

1.

Per for!ning a LCS redistribute mode functional test te verify the capability of rerouting most of the cooling water to the upper side valls and the top head.

2.

Performing a

functional test to verify the capability to increase the PCRV surge tank pre,sure to 30 psig by adding heliun to the tant,

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/cendment No.

Page 3/4 E-l l4Pf FEB 2 8 p-*u PLR(AND CONFINEMENT SYSTEMS 3/4.6.2 REACTOR _ PLANT COOLING WATEA/PCRV LINER COOLING SYSTEM -

SHUTDOWN LIMITING CONDITIONS FOR OPER/4TIONS 3.6.2.2 The Reactor Plarit Coeling Water (RPCW)/PCRV Liner Cooling System (LCS) shal) be OPERASLE with cne PPCW/PCRV l,0G loop OPERATING with at least one heat exchanger and cae pt.xp in each loop OPERATING.

APPLICABILITY: STARTUPa#, SHUT 00WN'#, and REFUELING #

ACTIDN: Vith no RPCW/PCRV LCS loop OPERATIWS, restore at least one

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loop to OPERATINS statos prior to reacAf qg a CALCULATED BUt.K CORE TEMPERATURE of 760 degrees F or suspend all operations involving CORE ALTERATIONS or control rod move:nent resulting in positive reactivity changes.

SURVEILUME fEQUIREMENTS

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4.6.2.2 IJo additional surveillance requiresents other.than those identified per Specification 4.6.2,1.

Whenever CALCULATED BULK CORE TEMEPRATURE is less than or equal to 760 degrees F.

The core sopport floor zone of the PCRV Liner Cooling System rnay be valved out when PCRV pressure is less than or equal to 150 psfa and CORE AVERAGE INLET TEMPEP.ATURE is less than or equa? to 200 degrees F.

FOR INFO I

ONLY

o Amendment No.

Page 3/4 6-DRAFT BASIS FOR SPECIFICATION LCO 3.6.2 / SR 4.6.2 O NO During operation at POWER, two PCRV liner cooling system loops are required to maintain PCRV liner cooling system temper.atures and stresses within the FSAR design limits (FSAR Section 5.9.2.,

Thermal Barrier and Liner Cooling Ssystem Design and Design Evaluation).

Analytical calculations in support of the PCRV Liner Cooling System design (FSAR Section 5.9.2.4) demonstrate that operation at full power with one cooling loop for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months satisfies the criterion which specifies a maximum temperature increase of 20 degrees F in the bulk temperature of the PCRV concrete.

Operation on one loop during a loss of forced circulation accident using a PCRV liner cooldown with an increased liner cooling water system cover pressure of 30 psig may result in temperature rises across individual cooling tubes of 240 degrees F (outlet temperature of approximately 340 degrees F). These onditions result in acceptable liner cooling for this analyzed condition and PCRV structural integrity is preserved (FSAR Section 0.1.2.1.5).

The liner cooling tubes are spaced in such a manner as to limit local concrete temperatures adjacent to the liner to 150 degrees F.

However, potential failures of cooling' tubes were analyzed and their limits follow.

PCRV liner cooling tube failures, whether the result of leakage or blocking, do not affect the integrity of the PCRV as long as such - #ailure is limited to a single tube in any adjacent set of four tubes on the PCRV cavity side walls, PCRV cavity bottom casing, core support floor side wall or core support floor liner bottom head, or a tingle tube in any adjacent set of six tubes on the PCRV cavity lirer top head and core support floor top casing.

A failed tube wnich doubles back on itself is considered a single tube failure.

In these cases, the local temperature in the concrete would be less than 250 degrees F (during normal two loop operation), an allowable and acceptable concrete temperature (FSAR 5.9.2.3.).

Operation of the PCRV liner cooling system during startup testing disclosed hot spots on the liner.

These locations were identified and analyzed in the above FSAR Sections.

The engineering evaluation indicated that operation with the hot spots would not compromise PCRV integrity and continued operation is acceptable. The temperature limits of the tubes associated with the hot spots are specified separately as they were analyzed specifically for each hot spot. Only foar of the s e'. e n hot spots have liner cooling tubes which may have temperature rises greater than 20 degrees F.

7

Amendmint No.

Page 3/4 6-DRAFT FEB 2 s 99 The ACTION times specified for recovery of two operating loops comes from analyses described in FSAR Section 5.9.2.4, i.e.,

48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months operation on one loop before temperature of.the bulk concrete would rise 20 degrees F.

With the number of: cooling tubes less than required, a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months action time is sufficient to identify and restore the tube to operating status (if possible) or SHUTDOWN to make permanent repairs.

The surveillance (s) and their respective intervals are specified to verify operability of the liner cooling system.

Components and features of the reactor plant cooling water system that are not safety related do not affect LCS operability.

The ISI/IST program at Fort St. Vrain verifies OPERABILITY of those barriers that separate safety and non-safety related portions of the systen.

A 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months surveillance on system flow rates provides additional verification of flow as process alarms monitor flow continuously in each liner cooling loop.

Individual tube failures would be expected to occur slowly, thus a 31 day SURVEILLANCE INTERVAL will detect tube failures'in time to take corrective action.

With CALCULATED BULK CORE TEMPERATURE less than or equal to 760 degrees F, one operating liner cooling system loop is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

When the PCRV pressure is less than 150 psia and CORE AVERAGE INLET TEMPERATURE is less than 200 degrees F, the core support floor zones of the liner coolfo1 system may be valved out as concrete temperatures will be less than the 250 degree FSAR limitation. Thus, leaking liner cooling tubes which are awaiting repairs will not contribute to potential moisture ingress into the primary system.

In Surveillance Requirement 4.6.2.1.b., tube outlet temperatures are determined by thermocouple readings.

In. the event of an instrument failure (i.e.,

a thermocouple is thought to be failed), the tube with the failed thermocouple may be considered OPERABLE if thermocouple readings for two adjacent tubes on either side of that tube are within their respective temperature limits.

If the tube itself failed rather than the thermocouple, then the temperature of adjacent tubes would be expected to rise.

Thus, a failed thermocouple can be identified vs an actual tube failure.

Power operation may continue until such time as the thermocouple can be repaired or replaced as long as the total of four adjacent tubes (two on either side of the tube with the fail _ed instrument) are within their respective temperature 1

limits.

Amendment N3.

Page 3/4 6-DRApr The use of 760 degrees F CALCULATED BULK CORE EMU @Ss a division between the APPLICABILITY of Specification 3.6.2.1 and 3.6.2.2 is explained as follows:

In the FSV HTGR, the limiting parameter of interest is a core inlet temperature greater than 760 degrees F.

The CALCULATED BULK CORE TEMPERATURE is a conservative calculation of the

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maximum potential temperature in the core and surrounding components.

The conservatisms are such that if the CALCULATED BULK CORE TEMPERATURE is limited to 760 degrees F,

the design inlet temperature of 760 degrees F is not exceeded. Systems used for accident prevention s.nd mitigation are required to satisfy the single failure criterion whenever CALCULATED BULK CORE TEMPERATURE is greater than 760 degrees F.

However, when CALCULATED BULK CORE TEMPERATURE is equal to or less than 760 degrees F, it is acceptable to require only one OPERABLE system required for accident prevention and mitigation as acceptable without single failure consideration, on the basis of the limited core cooling requitements.

All forced circulation may be interrupted for maintenance purposes provided that the time calculated for CALCULATED BULK CORE TEMPERATURE to reach 760 degrees F is not exceeded.

However, if forced circulation is temporarily restored, a

recalculation can be performed as required based on present plant conditions, to establish a new time period for CALCULATED BULK CORE TEMPERATURE to reach 760 degrees F.

Redundant systems may also be taken out of service for maintenance or surveillance testing provided that forced circulation is maintained. The time to reach CALCULATED BULK CORE TEMPERATURE equal to 760 degrees F may be recalculated as often as required.

Amendmont No.

Page 3/4 6-

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DRAFT PCRV AND CONFINEMENT SYSTEMS pgB2e1995 3/4.6.3 REACTOR PLANT COOLING WATER /PCRV LINER COOLING SYSTEM TEMPERATURES LIMITING CONDITIONS FOR OPERATION 3.6.3 The RPCW/PCP.V Liner Cooling System (LCS) temperatures shall be maintained within the following limits:

a.

The maximum average temperature difference between the common PCRV cooling water discharge temperature and the PCRV external concrete surface temperature shall not exceed 50 degrees F.

b.

The maximum PCRV Liner Cooling System water outlet temperature shall not exceed 120 degrees F.

c.

The maximum change of the weekly average PCRV concrete temperature shall not exceed 14 degrees F per week.

d.

The maximum temperature difference across the RPCW/PCRV Liner Cooling Water Heat Exchanger (LCS portion) shall not exceed 20 degrees F.

e.

The minimum average LCS water temperature shall be greater than or equal to 100 degrees F.

APPLICABILITY: At all times ACTION:

a.

With any of the above limits not satisfied, restore the limit (s) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or be in SHUTDOWN or REFUELING within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and suspend all operations involving CORE ALTERATIONS control rod movements resulting in positive reactivity changes, or movement of IRRADIATED FUEL.

Fog INFO ONLy

Am:ndment No.

.Page 3/4 6-DRAFT FEB 2'8 mes SURVEILLANCE REQUIREMENTS 4.6.3 The RPCW/PCRV Liner Cooling System temperatures shall.be demonstrated to be within their respective limits at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by:

a.

Verifying that the maximum temperature difference averaged over a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period between the PCRV external-concrete surface temperature and the common PCRV cooling water discharge temperature in each loop does not exceed 50 degrees F.

b.

Verifying that the maximum PCRV liner cooling water outlet temperature does not exceed 120 degrees F as measured by PCRV liner cooling water outlet tempereture in each loop.

c.

Verifying that the change in PCRV concrete temperature does not exceed 14 degrees F per week as indicated by the weekly average water temperature measured at the common 3

PCRV cooling water outlet temperature in each loop. The weekly average water temperature is determined by computing the arithmetical mean of 7 temperatures, representing each of the last 7 days of common PCRV cooling water outlet temperatures in each loop. Each day results in a new computation of a weekly average water temperature. The new weekly average is then compared to the weekly average water temperature computed 7 days earlier to verify the limit of Specification 3.6.3.c.

d.

Verifying that the maximum delta T across the RPCW/PCRV Liner Cooling System heat exchanger does not exceed 20 degrees F as-measured by the PCRV heat exchanger outlet temperature and the common PCRV liner cooling water outlet temperature in each loop.

e.

Verifying that the minimum average water temperature of the PCRV Liner Cooling System is greater than or equal to 100 degrees F as measured by the average of the PCRV Liner Cooling System heat exchanger (LCS side) inlet and outlet temperatures.

Amendment No.

Page 3/4 6-DRAFY FEB 2 e sea*

BASIS FOR SPECIFICATION LCO 3.6.3/ SR 4.6.3 The temperature limits associated with the Liner Cooli.ng System are not specifically discussed in the FSAR. Various FSAR sections-including 5.7, 5.9, 5.12, and 9.7 discuss general design limits of the liner and PCRV concrete. The PCRV liner and its associated cooling system assist in maintaining integrity of the PCRV concrete.

PCRV bulk concrete temperature is not measured directly. The PCRV Liner Cooling System temperatures and their specified frequency of measurement ensure that thermal stresses on the PCRV concrete and liner are within FSAR analyses described above and that PCRV integrity is maintained.

Since the PCRV concrete has a large thermal mass and inertia, temperatures would be expected to respond very slowly to any changes in the specified parameters. A 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months restoration and ACTION time is consistent with the expected slow temperature response of the PCRV. As a precaution, the plant would be SHUTDOWN and/or remain in REFUELING mode until temperatures were stabilized.

n

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Amendment Ns.

Paga 3/4.7-DRAFT PLANT AND SAFE SHUTDOWN COOLING SUPPORT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES - OPERATING LIMITING CONDITION FOR OPERATION 3.7.1.5 a. The steam generator superheater (EES) and reheater safety valves (V-2214, V-2215, V-2216, V-2245, V-2246, V-2247, V-2225 and V-2262). shall be OPERABLE with set points in accordance with Table 4.7.1-1, and

b. The provisions of Specification 3.0.6 are not applicable until 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after reaching 25% RATED THERMAL POWER, to allow testing of the steam generator superheater and reheater safety valves required following maintenance or per Surveillance Requirements identified in-Specification 4.7.1.5.

APPLICABILITY: POWER, LOW POWER, and STARTUP ACTION:

With one of the required safety valves inoperable, restore the required valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or restrict plant operation as follows:

1.

With an EES safety valve inoperable, reduce THERMAL POWER to less than 50% of RATED THERMAL POWER.

2.

With an EES safety valve inoperable while in STARTUP, restrict plant operation to a maximum of two boiler feed pumps.

3.

With a reheater safety valve inoperable, be in STARTUP within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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. Amendment No.

Paga 3/4.7-DRAFT SURVEILLANCE REQUIREMENTS FEB 2 8 mas

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4.7.1.5 The safety valves shall be demonstrated OPERABLE prior to exceeding 25% RATED THERMAL POWER. unless' completed in the previous five years by testing the'superheater and reheater safety valves as required by Specification 4.0.6, and by verifying the lift settings as specified in Table 4.7.1-1.

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Amendment No.

P ge 3/4.7--

DRAFT FE8 2 8199S 1i I.

TABLE 4.7.1-1

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C STEAM GENERATOR SAFETY VALVES VALVE NUMBER LIFT SETTINGS LOOP I V-2214 Less than or equal to 2917 psig V-2215 Less than or equal to 2846 psig-V-2216 Less than or equal to 2774 psig 4

V-2225 Less than or equal to 1133 psig LOOP II V-2245 Less than or equal to 2917 psig V-2246 Less than or equal to 2846 psic V-2247 Less than or equal to 2774 ps'.g V-2262 Less than or equal to 1133 psig j

1 a

4 4

a

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+,.,.-...--7

Amendment No.

Page 3/4.7-DRAFT PLANT AND SAFE ~ SHUTDOWN COOLING SUPPORT SYSTEMS FEB 2 8 GSE 3/4.7.1 TURBINE CYCLE SAFETY VALVES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.7.1.6 The steam generator superheater or reheater safety valve (s) which protect the OPELTING section(s) of the steam generator shall be OPERABLE with setpoints in accordance with Table 4.7.1-1.

APPLICABILITY: SHUTDOWN and REFUELING ACTION:

With less than the above required safety valve (s) OPERABLE, restore the required safety valve (s) to OPERABLE status prior to reaching a CALCULATED BULK CORE TEMPERATURE of 760 degrees F or suspend all operations involving CORE ALTERATIONS or control rod movements resulting in positive reactivity-changes.

SURVEILLANCE REQUIREMENTS 4.7.1.6 No additional surveillances required beyond those identified per Specification 4.7.1.5.

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Page 3/4.7-DRAFT FE8 2 81986 BASIS FOR SPECIFICATIONS LC0 3.7.1.5/SR 4.7.1.5 AND LCO 3.7.1.6/SR 4.7.1.6 The economizer-evaporator-superheater (EES) section of each steam generator loop is protected by three spring-loaded safety valves, each with one-third nominal relieving capacity of each loop. The reheater section of each steam generator loop is protected from overpressure transients by a single safety valve. These steam generator safety valves are described in the FSAR, Section 10.2.5.3.

The above valves are required to be tested in accordance with (ASME Section XI, IGV requirements) every 5 years or after maintenance. To satisfy the testing criteria, the valves must be tested with steam.

Since these valves are permanently installed in steam piping, the appropriate means for testing require plant power to be in excess of 22% RATED THERMAL POWER. Thus, the test must be conducted during LOW POWER. Conditions are specified so as to minimize operation at power until the valves are tested.

Due to the infrequent required testing of these valves, the likelihood of an accident occurring without proper valve testing is considered very small and plant safety is not compromised.

During all MODES, with one EES safety valve inoperable, plant operation is restricted to a condition for which the remaining safety valves have sufficient relieving capability to prevent overpressurization of any steam generator section (i.e.,

one boiler feed pump per operating loop).

Conversely, with any reheater safety valve inoperable, plant operation is restricted to a more restrictive Mode.

A 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action time for repair or SHUTDOWN due to inoperable safety valves ensures that these valves are returned to service in a

relatively short period of time, during which an overpressure transient is unlikely. Operation at poser for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months does not result in a significant loss of safety function for any extended period.

The setpcints for the safety valves identified in Table 4.7.1-1 are those valves identified in the FSAR with tolerances applied such that the Technical Specificatiens incorporate an upper bound setpoint. This is consistent with not incorporating normal operating limits in these Specifications.

ATTACHMENT 3 TO P-8616S COMMENTS ON NRC RESPONSE TO ACTION ITEMS i

d In Enclosure 3 to G-86008, the NRC requested that PSC resubmit responses to Action Items 27a, 27b, and 27c without adoption of the subjective s

criterion used by the Atomic Safety and Licensing Appeal Board in ALAB-

531, "that Technical Specifications are to be reserved for those matters i

as to which imposition of rigid conditions or limitations upon reactor operation is deemed necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety." (9 NRC 263, 1979)

The following is PSC's resubmittal of responses to those Action Items:

Action 27a PSC is to evaluate the acceptability of operation without buffer He as a circulator shaft seal (i.e., don't require buffer He flow in the Tech Specs).

Response

i Buffer helium is not required for continued operation of the helium circulators. However, the pressure boundary integrity of the buffer helium system is required to maintain primary coolant pressure in the bearing water surge tanks, insuring proper operation of the bearing water system.

If depressurization of both buffer helium loops occurs, cover gas on both bearing water surge tanks would be lost.

In this case, the backup bearing water syst(m insures circulator operation.

If the backup bearing water system is unavailable and PCRV pressure is greater than 300 psia, as discussed in FSAR Section 4.2.2.3.6, all three normal bearing water pumps in a loop can be operated to provide adequate bearing water flow to a circulator for safe shutdown cooling. Therefore, PSC proposes that buffer helium flow not be required in the Technical Specifications for helium circulator operability.

Action 27b PSC is to evaluate the need to specify maximum circulator bearing water temperature in the Tech Specs.

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Response

Bearing water temperature should not be specified for LC0 3/4.5.1.

The LC0 requirements and surveillances insure operabili.ty. of the bearing water system and FSV design features provide sufficient diversity and redundancy to insure sufficient cooling.

Loss of cooling water to the bearing water coolers results in an increase in.

water temperature tc the bearings.

Eventually, water temperatures would reach levels greater than 200 degrees F, but this would take about one hour. Subsequently, operation could continue by remote-manually switching to the backup bearing water supply which uses either service water, circulating water, or firewater as the coolant in the backup bearing water cooler.

Action 27c PSC is to evaluate the need to require a backup helium buffer gas supply to be specified in the Tech Specs.

Response

As noted in Action 27a above, neither buffer helium flow nor backup helium buffer gas supply is required for circulator operability.

Therefore, it is not necessary to specify either of these systems.

Action 27d Response - NRC Staff Comment:

The FSAR should be updated to clarify the licensee's position on circulator interlocks whose failure could prevent any source of motive power from being supplied to circulator drives.

Proposed Resolution PSC should submit this clarification in their next annual FSAR revision.

PSC Response Circulator interlock clarification will be provided in the next annual FSAR revision.

ATTACHMENT 4 TO P-86169 DISCUSSION ON CALCULATED BULK CORE TEMPERATURE The existing Technical Specifications for Fort St. Vrain generally require equipment to be operable "at power",

i.e., above 2 percent rated thermal power.

During the development of the upgraded Technical Specifications, in order to include the features of the Standard Technical Specifications (STS),

it was necessary to define equipment operabilities down to 0 percent power.

Requiring full redundancy in safe shut'down~ cooling systems at low power levels is not appropriate, considering the reduced decay heat lopds.

The STS do not require redundancy in ECCS systems below a reactor coolant system temperature of 350 degrees F, "on the basis of the stable reactivity condition of the reactor and the limited core cooling

~

requirements." (Westinghouse Bases, Rev 5,

3/4.5.2 and 3/4.5.3 ECCS Subsystems).

Consistent with this, PSC defined a Calculated Bulk Core Temperature, and established that as long as this parameter is below 760 degrees F, redundancy is not required. 760 degrees F was chosen as the most limiting design temperature in the PCRV.

This is a calculated average temperature of the core, including graphite and fuel, assuming a l'oss of all forced circulation of primary coolant.

Therefore, when Calculated Bulk Core Temperature is below 760 degrees F, the reactor is stable and core cooling requirements are reduced.

Figures 1 and 2 provide an indication of the time periods that are i

available after a loss of forced circulation, before the Calculated Bulk

~

Core Temperature reaches 760 degrees F.

Various core operating histories are included.

As noted in sections 3.1 and 3.6 of the FSAR, the expected gas coolant temperatures at full power are 760 degrees F (core inlet and upper plenum) and 1460 degrees F (core outlet and steam generator inlet) with a median fuel temperature of about 1500 degrees F.

The upper plenum internal components, control rod drive and orifice assembly and thermal barrier, have all been designed to be consistent with this temperature environment.

i 8

m - - - ---

In the systems description for the control rod drive and orificing assembly (50-12-1 through 5) it states that all of the materials used are consistent with a design gas temperature of 1150 degrees F.

In addition, in section 5.9 of the FSAR it states that the whole of the internal surface of the PCRV liner is covered by a Class A thermal barrier and that all of the metal parts of this barrier are dimensioned to acccmmodate thermal expansion clearances to 1000 degrees F.

Consequently, limiting the Calculated Bulk Core Temperature during a primary coolant flow termination to 760 degrees F would conservatively ensure that both the core and PCRV internals would be protected when the primary coolant flow is resumed. This is consistent with the conclusion reached by ORNL in their independent review (FIN No A9351).

Calculational Method PSC presently calculates a bulk core temperature for the purpose of determining how long forced primary coolant flow may be suspended via procedure CMG-4, attached.

Prior to implementation of the upgraded Technical Specifications, this procedure will be revised to reflect the Calculated Bulk Core Temperature terminology, but the methodologies will be essentially the same.

Administrative Controls Decay heat calculations are currently performed by Technical Services engineers as required. Prior to implementing the upgraded Technical Specifications, PSC intends to make the Calculated Bulk Core Temperature readily available to the operators.

At the present'

time, PSC anticipates that the plant computer will be the most likely source for this parameter.

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ML20137X211

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