ML20209B721

Forwards Comments on Tech Spec Upgrade Program for Limiting Condition for Operations (Lcos) for safety-related Cooling functions,marked-up Redrafted LCOs & List of NRC Comments Addressed in Encl 1,for Review within 75 Days of Ltr Date

ML20209B721
Person / Time
Site:	Fort Saint Vrain
Issue date:	04/17/1987
From:	Heitner K Office of Nuclear Reactor Regulation
To:	Robert Williams PUBLIC SERVICE CO. OF COLORADO
References
NUDOCS 8704280502
Download: ML20209B721 (60)

Text

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/ UNITED STATES y 1.> q NUCLEAR REGULATORY COMMISSION

.j W ASHING TON. O. C. 20555 o, a

~*' . . . . . **' M R n 1997 Docket No. 50-267 Mr. R. O. Williams, Jr.

Vice President, Nuclear Operations Public Service Company of Colorado P. O. Box 840 Denver, Colorado 80201-0840

SUBJECT:

NRC COMMENTS ON THE TECHNICAL SPECIFICATION UPGRADE PROGRAM (TSUP),

LCOs FOR SAFETY-RELATED COOLING FUNCTIONS

References:

(a) H.L. Brey letter to H.N. Berkow, Technical Specification Upgrade Program, Redrafted LCOs, February 28, 1986, Public Service Company of Colorado.

(b) 0.R. Lee letter to H.N. Berkow, Final Draft of Upgrade Technical Specifications, November 27, 1985, Public Service Company of Colorado.-

(c) K.L. Heitner letter to R. F. Walker, NRC Comments on the Final Draft of the Fort St. Vrain (FSV) Upgraded Technical Specifications, May 30, 1986, U.S. Nuclear Regulatory Comission.

Enclosure 1 forwards our coments on the TSUP final draft LCOs for systems, sub-systems and components with safety-related cooling functions. Enclosure 1 provides both comments on the redrafted LCOs certified in Reference (a) and additional connents on selected definitions, sections and LCOs of the TSUP final draft identified in Reference (b). The additional cornrents on the previously reviewed portion of the TSUP final draft are necessary to clarify the safety-related cooling functions which are actually relied upon in the Fort St. Vrain licensing basis as embodied in your updated FSAR. The need for the additioral clarifications became evident during a comprehensive review of the redraftt.d LCOs identified in Reference (a). This review addressed equipment redundancy necessary during normal and abnormal operation to assure cooling system availability for the spectrum of events analyzed in your updated FSAR. The additional comments supplement, but do not replace, the previous NRC comments identified in Reference (c).

Enclosure 2 provides a markup of the redrafted LCOs that were identified in Reference (a). Enclosure 3 tabulates the list of NRC comments addressed in Enclosure 1 and further correlates each ccmment with a comment category designation using the category identifiers which were agreed upon in our meetings with your staff on Octcber 1 and 2,1986, and again on October 27

' through 30, 1986. The specific comments categories utilized in Enclosure 3

.are defined as follows:

8704280502 870417 i

PDR ADOCK 05000267 P PDR

f Mr. R. O. Williams ,

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B - No further cction is required, D - Further PSC/NRC discussion is needed to resolve connent, D* - NRC action is required, 3

F - Outside TSUP scope, but further discussion may be needed.

Except for the evaluation of your proposed PRA for DBA-2, this completes our review / comment of Reference (a). Therefore, we request that you review Enclosure 1, 2 and 3 and provide your comments within 75 days of the date of this letter.

Should you have any questions'regarding this letter and the enclosed comments, please contact me at (301) 492-8205.

The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511.

Sincerely, Kenneth L. Heitner, Project Manager Standardization and Special Projects Directorate Division of PWR Licensing-B Office of Nuclear Reactor Regulation

Enclosures:

1. NRC Comments on TSUP Draft Specifications for Systems, Subsystems and Components with Safety-Related Cooling Functions.

2. NRC Markup of draft LCOs from the PSC submittals dated February 28, 1986.

3. Categorization of NRC Comments.

cc w/ enclosures:

See next page DISTRIBUTION:

Docket File .FMiraglia JPartlow KHeitner NRC PDR OGC-BETH HThompson 0 Lynch Local PDR EJordan ACRS(10) HBerkow PBSS Reading BGrimes PNoonan J C ^ W4 PJ h nan PBSS M d KHeitner:cw PBSS-Olynch PBSS4 H8erkow Pbt JA.Gilvo 03/9 /87 03/10/87 03/-/87 03/- -/87 t/ /g// 7 i

t, ,

Mr. R. O. Williams Public Service Company of Colorado Fort St. Vrain l cc:

Mr. D. W. Warembourg, Manager Albert J. Hazie, Director

Nuclear Engineering Division Radiation Control Division

Public Service Company Department of Health

• of Colorado 4210 East 11th Avenue .

P. O. Box 840 Denver, Colorado 80220 Denver, Colorado 80201 Mr. David Alberstein, 14/159A Mr. R. O. Williams, Acting Manager i GA Technologies, Inc. Nuclear Production Division Post Office Box 85608 Public Service Company of Colorado San Diego, California 92138 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Mr. H. L. Brey, Manager Nuclear Licensing and Fuel Division Mr. P. F. Tomlinson, Manager Public Service Company of Colorado Quality Assurance Division P. O. Box 840 Pubite Service Company of Colorado Denver, Colorado 80201 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Senior Resident Inspector

. U.S. Nuclear Regulatory Comission Mr. R. F. Walker i P. 0. Box 840 Public Service Company of Colorado

Platteville, Colorado 80651 Post Office Box 840

. Denver, Colorado 80201-0840 i

Kelley, Stansfield 8 0'Donnell Public Service Company Building Comitment Control Program Room 900 Coordinator 550 15th Street Public Service Company of Colorado j Cenver, Colorado 80202 2420 W. 26th Ave. Suite 100-0 Denver, Colorado 80211

Reofonal Administrator, Region IV i

U.S. Nuclear Regulatory Commission )

611 Ryan Plaza Drive, Suite 1000 '

Arlington, Texas 76011 Chainnan, Board of County Comissioners of Weld County, Colorado Greeley, Colorado 80631 1 j Pegional Representative

, Radiation Programs Environmental Protection Agency 1 Denver Place 999 18th Street Suite 1300 j Denver, Colorado 80202-2413 i

4

p Enclosure 1 1

NRC Comments on TSUP Draft Specifications for Systems, Subsystems, and Components with Safety-Related Cooling Functions NRC COMMENT - DEFINITION 1 34 (NRC Markup of Final Draft)

PSC needs to be definitive and specific with regard to the use of Safe Shutdown Equipment referred to in this definition and with r.egard to what constitutes the list of Safe Shutdown Equipment per the FSAR. PSC needs to clarify these points within DEFINITION 1.34 (in the NRC markup of final draf t) . The current wording of this definition implies that safe shutdown equipment includes whatever PSC so' chooses to include as such in TSUP draf t Section 3/4.5. Therefore, this definition needs to be modified to be consistent with the FSAR Section 1.4 definition of Safe Shutdown Equipment.

However, PSC further needs to clarify the FSAR as to (1) how items listed on FSAR Tables 1.4 1 and 1.4 2 and in FSAR Figure 10.3 4 and those cited in FSAR Sec,tions 1.4.5.3 and 10.3.9 compare with each other.

(2) how the same set of items compares with the " listing of the Safe Shutdown Cooling Equipment ... maintained at the Fort St. Vrain site" per FSAR Section 1.4.5.4 and (3) how the combined sets compare with the list of " Fort St. Vrain's Class I and Safe Shutdown electrical equip-ment" per FSAR Section 1.4.6. PSC needs to differentiate clearly among

" safe shutdown," " safe shutdown cooling," " emergency cooling," and

" essential cooling" equipment! Apparently, there exists " essential ,

cooling" equipment not on the " safe shutdown" list and which PSC does not intend to subject to Technical Specifications.

If the FSAR defines a piece of equipment as " essential." then it follows that that equipment should be subjected to a measure of compliance against its " essential" function. Are there different lists as implied by the FSAR? What are the differences bet een the Safe Shutdown Equipment list and the Safe Shutdown Cooling Equipment list? Which list contains the complete set of Safe Shutdown electrical equipment?

Further, the " Remarks" columns in FSAR Tables 8.2 4 through 8.2 7 should be revised and clarified to denote whether, for each " essential load" of the standby diesel generators, the identified component (1) is on the Safe Shutdown (Cooling?) Equipment lists, (2) is an " emergency cooling" component required per FSAR accident analysis (for example, DBA 2), but is not otherwise listed elsewhere as noted above, or (3) is an "essen-tial" component relied upon to effect " normal shutdown cooling" or

" residual heat removal" following normal shutdown as well as following upsets such as loss of offsite power.

Finally, the list of ACM diesel generator loads (FSAR Table 8.2 9) should perhaps be ordered by the procedural start sequence, page 8.2-32ff, in a manner similar to FSAR Tables 8.2 4 through 8.2 7 and should in such case incorporate appropriate remarks as to the equipment qualif.

ication status or other des 1 6nation, as noted above.

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' 2 As the FSAR is currently written, there are either too many equipment lists or perhaps too many distinctive yet unclear designations for the same equipment list. Unique equipment designations should be defined clearly and completely with regard to qualification requirements. Com-ponents appearing on more than one list should be clearly identified and c ro s s - re fe rence d. The Specifications should be revised accordingly to reflect the proper equipment designations. The attached table is pro-vided as a guide only to demonstrate in a simple fashion NRC's current understanding of the Fort St. Vrain. abnormal shutdown cooling systems.

The attached table includes acknowledgment of the 10 CFR 50, Appendix R, requirements for which no Technical Specifications are yet defined.

PROPOSED RESOLUTION PSC needs to revise the DEFINITION and to clarify the FSAR as indicated in the comment. The attached table should be used as a guide in defin-ing abnormal shutdown cooling systems requirements for which additional Technical Specifications may be required.

NRC COMMENTS - Table 1.1 (NRC Markup of Final Draft)

1. PSC has chosen to use the value of the CALCULATED BULK CORE TEMPERA-TURE within the TSUP draf t to demarcate the need for redundancy require-ments among safety-related cooling components within the SHUTDOWN mode.

The process of making distinctions between the operating modes would appear to be better served by using the limiting value of the CALCULATED CORE BULK TEMPERATURE, namely, above or below 7600F, to mark the transi-tion from HOT SHUTDOWN to COLD SHUTDOWN similar to the W STS. This demarcation should perhaps also be made contingent upon PCRV pressure, such as above or below 100 psia.

2. In LCO 3.9.1, the REFUELING mode is further defined by the condition of having the CORE AVERAGE INLET TEMPERATURE below 1650F. Why is this coniition not given in Table 1.1 so as to be readily apparent as also applying to all LCOs with REFUELING listed in the APPLICABILITY state-ment? Also, what is the value of the CALCULATED BULK CORE TEMPERATURE supposed to be during REFUELINGt Such APPLICABILITY conditions should be made explicit and direct.

PROPOSED RESOLUTIONS 1.

PSC needs to explain the advantages of not incorporating the sug-gested changes.

2. PSC needs to provide changes to assure clarity in the TSUP draft.

NRC COMMENTS Section 3/4.2

1. TSUP final draft Specifications 3/4.2.2, 3/4.2.3, and 3/4.2.4 address the methods used by the operator to demonstrate compliance with limits imposed upon regionwise radial power peaking. The assumptions about intra region or local power peaking and axial power peaking are addressed briefly in DESIGN FEATURE 5.3.4 and more specifically in the

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, ' 3 Table Simplified Breakdown of Fort St. Vrain (FSV) Abnormal Shutdown Coolinc FSV Power Sources OP - Offsite Power EP - Emergency (Standby) Diesel Cenerators (Class LE)

ACM - Alternate Cooling Method Diesel Cenerator (Non Class lE)

Maior FSV Heat Removal Systems or Components BFP - Boiler Feed Pump (Non Class I)

CS - Condensate System (Non Class I)

FS - Firewater System (Class I)

RPCW/LCS - Reactor Plant Cooling Water / Liner Cooling System (Class I)

LWR Equivalent Cooling Functions AFW/RER - Auxiliary Feedwater/ Residual Heat Removal CHR - Containment Heat Remqval ECC - Emergency Core Cooling FSV Abnormal Shutdown Cooline Scenarios Scanario Power Heat Removal LVR Ecuivslent Function Seismic # (Class 1E) EP FS ECC Non Seismic (Class 1E) EP CS AFW/RHR 10 CFR 50, Appendix R (Fire)

ConSested Cable Areas ACM RPCW/LCS CHR Non. Congested Cable Areas EP or ACM CS or FS AFW/RHR or ECC DBA 1" EP RPCW/LCS CHR d

DBA 2 OP BFP ECC SAFE SHUTDOWN COOLING for the Design Basis Earthquake or Maximum Tornado, or for other events resulting in up to 90 minute interruption of forced cooling.

b Such as loss of offsite power and trip of the main turbine-generator set.

" Permanent Loss of Forced Cooling.

d Design Basis Depressurization.

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4 i Basis for SAFETY LIMIT 2.1.1, which in turn is satisfied by compliance

with Specification 3/4.2.6. However, compliance with limits on local,

or intra region, power peaking and axial power peaking as specified in

, FSAR Sections 3.5.4.2 and 3.5.4.3 is not based on an LCO.' Apparently, there are calculations but no checks nor surveillance requirements.

Such calculations, their associated quality assured verification, '

i including uncertainty analysis, and the checks for compliance against

local and axial power peaking limits are apparently performed outside i -

the purview of the operators. Therefore, this information could only come under the purview of the operators by providing the operators with -

the documentation of these calculations, their verifications and compli.

I ance checks. In this circumstance the operator is limited to comparing

such documentation against the basis for compliance as documented both

, within the FSAR and within the bases for any associated Specification, 1 Safety Limit or Design Feature.

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. Currently, power peaking and power peaking uneartainties, as presented j in FSAR Sections 3.5.4, 3.6.4, 14.2, 14.4, and D.1.2.1.4, are either  ;

j inadequately or simply not correlated with each other within the FSAR.

i Further, the FSAR also has documented no rigorous verification or uncer-tainty analysis of the Fort St. Vrain calculational methods compared to l

experimental data for power peaking. Such verification and uncertainty l .

analysis is expected to have been included in the Fort St. Vrain FSAR j

because the FSAR was apparently written under the AEC DRL guideline as provided in A Guide for the Organization and Contents of Safety Analysis

! Reports, TID-24631, June 30, 1966. TID 24631 provided the AEC DRL's l guidelines for demonstrating compliance with 10 CFR 50.34; and TID 24631 I is the predecessor to the current NRC Regulatory Guide 1.70. (The equivalent HTCR Edition for Regulatory Guide 1.70 is given in WASH 1266, July 1973). Per TID 24631, the presentation of the nuclear design in Section 3 of the safety analysis should include the " gross and local l '

1 radial and axial power distribution for different planned rod patterns" j

and for different xenon / samarium conditions, and "the nuclear evaluation

[in Section 3 of the safety analysis] should include a description of ,

! the calculational methods employed in arriving at important nuclear

[ parameters with an estimate of accuracy by comparison with experiments '

or with the performance of other reactors (underlining for highlight]J"

} Section 4.3 of the current regulatory guidance (WASH 1266) for HTCRs is j even more explicit with regard to documenting experimentally derived i power peaking uncertainties.

}

In Appendix III.3. A, Attachment A to Amendment No. 3 to the PSAR

] (DOCKET 50267 13), dated July 1967, PSC specifically committed to "the j analysis of the following experiments (namely, the Peach Bottom critical i experiments and the Peach Bottom initial core loadings) (which) will j allow acceptable estimates of the accuracy and uncertainty in the PSC j nuclear design to be made." In the same document, PSC also made interim j power peaking uncertainty estimates from Peach Bottom critical experi-l J ment correlations and acknowledged that there was " uncertainty in flux

) peaking prediction" due to " difficulty in showing accurately the actual  !

{ core geometry and composition in the calculational model." It is l'

, observed that PSC's qualitative or intuitive estimates of 54 and 106 i

j uncertainties in gross and local power peaking respectively, as ,

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5 documented in the PSAR amendment, do not appear to be well supported by comparisons made in Figure 12 of CA-3799, March 5, 1963 It is noted that an intuitive estimate of uncer.tainty might be made of at least twice that made by PSC from the same data. The appropriateness of such estimates is questioned given the assumed differences between models and cross section data used in 1963 and those used today for Fort St. Vrain.

It is further observed that on page III.3. A 4 of the PSAR amandment, PSC implied that, although the " accuracy of PSC design methods" based on such comparisons were felt by TSC to be adequate. PSC apparently commit-ted to using "new cross section data and larger and faster computer codes" to demonstrate that "this accuracy will improve." No results of i

either PSC's stated or PSC's implied commitments with regard to power peaking uncertainty are currently documented in the FSAR. Again such comparisons would be expected per the applicable guidance of TID-24631 and based on PSC's apparent commitments to do so.

Without an adequate, well documented licensing basis for verifying com-pliance with limits for local and axial power peaking, the bases for the above listed TSUP draft Specifications do not provide the operator with

' the resources to ascertain in a reasonable and unqualified manner whether local power peaking is acceptable for observed plant conditions.

In the STS, Specifications are provided that allow the operator to make i

such a determination directly from measurable reactor quantities, for example, incore flux maps. Since, unlike the STS, the TSUP Specifica-tions will not or cannot provide for a direct index of compliance, the licensing basis must be sufficiently clear as to why the index of com-

• pliance either is already establishest or does not require further specification. To achieve this, the licensing basis as embodied within the FSAR should be clarified in acecrdance with both the AEC-DRL gui-dance for content and organization of the FSAR and PSC's apparent previ-ous commitments to do so. In the absence of a Technical Specification the operator should be able to refer to the FSAR for a clear resolution of all questions re6arding local and axial power peaking and the associ-ated uncertainties.

2. Recent interactions between NRC and PSC with regard to the Fort St, Vrain Fuel Surveillance Program has indicated that the inclusion of gamma scanning of exposed fuel elements under this program has been made contingent by PSC upon the availability of DOE funding. To date. PSC's submittal of gamma scanning results has merely endorsed the CA analyses

' of these results to NRC without PSC comment (see PSC letter P 84532, dated December 17, 1984, and forwarding CA document 907079, Postirradiation Examination and Evaluation of FSV Element (1 2415)). On page 20 of the cited CA report, the claim is made that the " data (inferred power distribution) was used to verify nuclear design calcula-tions." On page 35 of the cited CA report, a disclaimer reads that

" verification of HTCR core design methods cannot be accomplished from comparisons of experimental observations and design code calculations for one element." This disclaimer was to be applied to a subsequent set of comparisons documented in the report. The reported comparisons on pages 36 and 37 begin with measured, b'!t not calculated, radial power distributions, and so the disclaimer apparently did not apply there, on page 37, calculated results for axial power peakin6 were reported

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apparently based on synthesized results for the fuel block. The differ-i ence between measured and calculated axial power tilting was 9%. No where in the GA report are comparisons made with FSAR limits or FSAR assumptions on local or axial power peaking. Does PSC endorse these comparisons as a legitimate verification of the Fort St. Vrain calcula-tional methods for predicting local power peaking?

In the published proceedings of Cas-Cooled Reactors Today, Volumde 3, 1982, analysts from the British Central Electricity Generating Board presented a paper entitled "The Role of Post Irradiation Examination of CACR Fuel in Evaluating and Improving Prediction of Power Distribu-tions," Paper No. 136. The referenced paper portrays a significantly more comprehensive use of gamma scannin5 for quantifying power peaking and power distribution uncertainties for the Advanced Cas-Cooled Reac-tors than has been presented elsewhere for Fort St. Vrain. It is noted that to date NRC's consideration of the use of gamma scannin6 for Fort St. Vrain design verification has apparently been limited to assuring 4

materials and structural mechanical performance of individual fuel i

blocks as opposed to the detailed verification of local radial and axial power peaking predictions by Fort St. Vrain neutronic design and analysis models and methods. Does PSC have a definitive, quantitative method of demonstrating continued compliance with assumed local and axial power peaking limits as documented in the licensing basis?

PROPOSED RESOLUTIONS

1. PSC needs to provide an LCO on local and axial power peaking. As a i

possible alternative PSC may clarify the licensing basis with regard to limits on axial and local radial power peaking, the associated power peaking uncertainties, and the mechanism for compliance with the speci-fled limits based on applying verified and quality assured analytical methods. In the latter case, the FSAR and the Bases for Specifications 3/4.2.2, 3/4.2.3, 3/4.2.4, and 3/4.2.6 would have to be revised as necessary to provide the operator with clear and complete information on the status of axial and local radial power peaking. Clarification to the FSAR should be made in accordance with the applicable regulatory guidance under which the FSAR was prepared and should be consistent with the objectives of previous PSC commitments with regard to providing

" acceptable estimates of accuracy and uncertainty."

2. PSC needs to provide specific reasons why an additional Specifica.

tion or an expanded Fuel Surveillance Program is not required to demon-strate compliance with FSAR limits on axial and local radial power peak-ing. In addition, PSC needs to provide clarifications to the questions noted within the comment, NRC C0KMENTS Section 3/4,4 i

1. On page 1, Section B, Attachment I to P 86169, PSC justified not incorporating "a requirement for OPERATING primary coolant loops, with j

numbers of OPERATINC helium circulators corresponding to various reactor power levels," by indicating that such a requirement "It not essential in the Technical Specifications, as it does not reflect any FSAR 4

' 7 requirement." However, FSAR Sections 4.3.1, 4.3.4 and 7.1.2.6 indicate that the nonsafety plant regulating system is relied upon to control helium circulator speed and to assure, at a minimus. power to-flow equivalency during abnormal upsets involving loss of circulators or reductions in the circulator rotating speed. Thus, the nonsafety plant regulating system is apparently committed to fulfi' ling an FSAR require-ment when the circulators are operating on steam drive, but the FSAR does not analyze situations involving failures of the plant regulatirg system to demonstrate meeting this FSAR requirement.

Also, FSAR Section 10.2 indicates that at least one of the auxiliary boilers must be used to effect normal cooling during reactor startup or shutdown. Auxiliary boiler steam is used to drive the helium circula-tors and/or the turbine driven boiler feed pumps. FSAR Section 10.2 is unclear about whether reactor startup actually involves the use of feed-water for circulator turbine drives under conditions of fission heating; however, the last paragraph of FSAR Section 10.2, page 10.2 2, as well as FSAR Sections 10.3 and 14.4 imply that such configurations are used to effect both orderly and abnormal shutdown of the reactor plant.

Thus, the use of feedwater as the circulator drive could presumably be interpreted to include portions of the shutdown maneuvers with fission heat still being generated in the core.

The FSAR is totally silent on the subject of whether fission heat is i

allowed to be generated when both the circulator drives and steam gen-erators are supplied only by condensate. Powered operation on condon-sate even during STARTUP (<5% rated) would not appear to be advisable given the limited coolant circulation capability under such conditions.

Does PSC currently or has PSC in the past allowed such operation?

As opposed to PSC's position in Attachment 1 to P 36169, the FSAR indi.

cates that the continuation of thermal margins to fuel damage, both dur.

ing normal powered operations and during startup and shutdown maneuvers, relies upon achieving a power to flow ratio with a value less than or equal to unity in order to assure adequate removal of fission heat from the core. The fact that the FSAR is so written as to obscure the pre-ferred or allowable methods of assuring thermal margins should not be i

used to avoid implementing Technical Specifications. The W STS, for example, provides direct surveillance of reactor core flow per W STS Specification 3/4.2.2 but still provides for assurance of an adequato configuration to assure flow per W STS Specification 3/4.4.1.1. The TSUP draf t essentially provides for the operator to attempt to adjust the core flow orifices without first assuring the adequacy of the integral flow.

This is inconsistent with both the FSAR and the W STS, Also, per ALAB 531, which was previously cited by PSC, Technical Specif-ications are "doemed necessary to obviate the possibility of an abnormal situation or event giVing rise to an immodlate threat to the public health and safety." Until objective criteria are developed by the NRC and given that PSC has based the TSUP on an underlying "immediate threat" philosophy, this statement appears to be best construed to mean

that the Technical Specifications should aid in preventing the inception of abnormal situations and events or the degeneration of such situations and events into potential threats to public health and safety. As such, 1

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8 the timely assurance of adequate core flow and steam generator cooling is judged necessary to prevent the entry into or the complication of an abnormal cooling configuration. A requirement for specifying OPERATINC '

primary coolant loops, including the use of auxiliary water and steam drive sources for powering the helium circulators, during fission heat-ing, is both appropriate and necessary.

2. On page 2 Section E Attachment 1 to P 86169, PSC contends that

" condensate drive for the Pelton wheels is a design feature of FSV that is not relied upon in the safety analysis" and that " firewater is the only Class I safety related Pelton wheel drive source relied upon for accident analysis." However, FSAR Tables 8.2 4 through 8.2 7 and FSAR Sections 10.2 and 10.3 indicate that the condensate drive for the Pelton wheels and the condensate source for the steam generators constitute the equivalent to the W-STS residual heat removal (RHR) loops and auxiliary feedwater (AFW) system. The purpose of these W-STS systems is to pro-vide normal shutdown cooling for the continuation of core thermal mar-gins against decay heat loads without being forced to rely on the safety grade emergency core cooling system. Therefore, the condensate drive of the Pelton wheels has a demonstrated equivalent safety function to that exemplified within the W STS and is so relied upon in the FSV safety analysis under both normal and accident conditions, including loss of offsite power with concurrent turbine trip. The fact that at FSV only the " untreated" firewater system is Class I is perhaps an unfortunate evolutionary development particularly since the FSAR desig-nates the two 12.54 capacity condensate pumps as " essential" loads for the diesel generators. It is noted that not all portions of the RHR and AFW systems are Class I rated at PVR plants and that there is variation in Class I components from plant to plant, but the PVRs do appear to have more Class I components associated with cleaner sources of water than FSV. The reliance on the firewater system at FSV to effect emer-gency core cooling may well assure safety in offsetting the effects of design basis events, but the poor water quality may be disastrous for potentially restart and operation. Considerations relative to the abil-Lty to restart and operate the plant af ter an accident could make the l

use of firewater a less than desirable "last resort."

3.

FSAR Tables 8.2 6 and 8.2 7 and FSAR Sections 10.3.1 and 10.3.5 indicate that the two 74 capacity circulating water pumps are relied upon to insure residual heat removal (RHR) during upset conditions, such i

as loss of offsite power with turbine trip, as well as to provide redun-l dancy during normal operation. Those pumps are an " essential" load on the diesel generator 'B' sequence for automatic start. FSAR Section 10.3.2 indicates that the condensate storage tank and the reheat steam relief valves provide the necessary backup to the operation of the 74 capacity circulating water pumps if the 'B' sequence of " essential" electrical loads cannot be accomodated during loss of offsite power with turbine trip and with the unavailability of one standby diesel Eenera-tor.

In FSAR Table 8.2 6, Sequence Load Number 5, the word " circulator" should be spelled " circulating."

4 FSAR Sections 10.2 and 10.3.1 indicate that when the core decay heat load falls to about it of rated reactor power, the decay heat removal

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• 9 exchanger is used at Fort St. Vrain to provide the equivalent function to that of the residual heat removal exchanger on the RHR loop of a LVR.

The RHR loop on an LWR is subject to OPERABILITY conditions within the W.STS during HOT SHUTDOWN, COLL SHUTDOWN, and REFUELING. Is the Fort St. Vrain decay heat removal exchanger used during REFUELING to maintain the CORE AVERACE INLET TEMPERATURE at or below 1650F per the require-ments of TSUP Specification 3/4.9.1 and FSAR Section 9.1.1.47

5. If PSC intends for the auxiliary boiler feed pumps to be considered as backup to the condensate pumps (per FSAR Section 10.3.6 and FSAR

• Appendix C.48) for RHR, can PSC quantify the " reduced condensate flow" (per page 4, SD 841) and its effect on primary coolant flow by condensate-driven circulators? FSAR Section 14.4.2.1 states that "one helium circulator can provide nearly 4.5% of rated flow through the reactor core when operating by itself with condensate water supplied to its water turbine drive." What is the rated primary system flow with

" reduced condensate flow" via the auxiliary boiler feed pumps during abnormal RHR configurations? Would the emergency water booster pumps be used to enhance condensate flow in any circumstances?

6. TSUP draft Section 3/4.4 should be retitled PRIMARY COOLANT SYSTEM.

PROPOSED RESOLUTIONS

1. PSC should provide LCOs for the required number of OPERABLE helium circulators on steam drive and feedwater drive as a function of power level in the POWER, LOW POWER and STARTUP modes of operation. PSC should provide LCOs for the OPERATING and OPERABILITY requirements for the auxiliary boilers during LOW POWER and STARTUP as needed to supply

• the circulator steam drives and boiler feed pump turbines to assure normal cooling during startup and shutdown maneuvers. PSC should revise the FSAR to clarify that the condensate drive will not be used as a nor-mal or routine method for removing direct fission generated heat unless a special exception is requested and approved by NRC.

2. PSC needs to clarify the FSAR with regard to the indicated safety function, namely, residual heat removal, provided by the condensate sup-ply to circulator water turbines and steam generators. PSC needs to confirm how potential operator reluctance to use untreated firewater for emergency core cooling will not affect timely operator responsed to situations in which emergency core cooling is required. Further, PSC needs to provide LCOs for the OPERABILITY and OPERATING conditions of the 12.5% capacity condensate pumps as part of the effective or equivalent residual heat removal (RHR) and auxiliary feedwater (AFW) systems at Fort St. Vrain.

3./4 PSC needs to provide LCOs for the effective or equivalent resi-dual heat removal (RHR) system at Fort St. Vrain. PSC needs also to respond to the question in Comment 4 Further, PSC needs to correct the misspelling in FSAR Table 8.2 6.

5. PSC needs to ptovide clarifications only if the auxiliary boiler fond pumps are to ao credited as being redundant backup to the

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, 10 condensate pumps and particularly if such a redundancy argument is to be used to justify no need for RHR Technical Specifications. Similarly, the potential condensate flow via the emergency water booster pumps would require quantification and detailed discussion if the need for RHR Technical Specifications is to be discounted by PSC based on the availa-bility of timely and redundant sources of condensate flow to accomplish RHR.

6. PSC needs to retitle the TSUP draft Section 3/4.4 to be consistent with the STS since the recommended new LCOs, which are derived from the Fort St. Vrain licensing basis, expand the scope of this section beyond that of chemical and radioactive impurities in the primary coolant.

NRC COMMENT - Section 3/4.5 PSC needs to retitle TSUP draft Section 3/4.5 to EMERGENCY CORE COOLING SYSTEMS, PSC needs to relocate into this section those Specifications which are currently located in TSUP draf t Section 3/4.7 and which address systems, subsystems, and components that must operate to effect emergency core cooling under accident conditions identified within the FSAR even though such equipment is not qualified as SAFE SHUTDOWN COOL-ING equipment per FSAR Section 1.4 Within Section 3/4.5, PSC needs to delineate and differentiate between those systems, subsystems, and com-ponents which are qualified as SAFE SHUTDOWN, or SAFE SHUTDOWN COOLING, and those which are relied upon for emergency core cooling without specific qualification (such as DBA 2 cooling). Please refer to the NRC comments on LCOs 3.7.1.1, 3.7.1.2, and 3.7.3 for the additional reasons for relocating Specifications to Seciton 3/4.5.

PROPOSED RESOLUTION PSC needs to revise the TSUP draft as indicated.

NRC COMMENTS . LCO 3.5.1.1

• 1. In LCO 3.5.1.1.a.2, the words " including two OPERABLE flow paths" should be inserted after the last word "0PERABLE" in the condition statement for the operability of steam generator sections. This change is consistent with stated conditions for OPERABILITY as given in the Basis for LCO 3.5.1.1 on page five under the section entitled Steam Cenerators.

2. In LCO 3.5.1.1.b.1, why is the term " safe shutdown cooling drive" used when the only sources of motive power for 8000 rpm circulator speed at atmospheric pressure are the boiler feed pumps? The boiler feed pumps are not included on the SAFE SHUTDOWN COOLING equipment list in either FSAR Table 1.4 2 or FSAR F16 ure 10.3 4 Should et.) circulator turbine drive requirements for DBA.2 cooling be addressed in a separato LC07 Should not the DBA 2 cooling system bo desi6nated as a separate category of equipment?

• 3. In LCO 3.5.1.1.b.2, the words "Two safo shutdown coolin8 drives" would appear to read more appropriately as "A safe shutdown cooling

• , ,

• 11 drive." There is only one steam and one water turbine drive per each circulator, and the text should reflect one drive oer OPERABLE circula-tor since the steam drive is not a safe shutdown c'ooli.ng drive.

• 4 In LCO 3.5.1.1.b.4, the words " normal bearing water system" are understood to refer to the full complement of three bearing water pumps per loop. The backup bearing water supply system is asrumed by the Staff neither to be credited in the FSAR nor subject to Technical

~

Specifications. PSC's revised response to Action 27a, as documented in Attachment 3 to P 86169, is interpreted to mean that backup bearing water is not required for effecting " safe shutdown cooling." Similarly, the PSC response to Action 27b is not appropriate since backup bearing water is neither covered by the Technical Specification nor available (in use) below 30% of rated power.

• 5. Section 4.0 and LCO 4.2.1 in the current FSV Technical Specifica-tions allow only a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period for an orderly shutdown when a loop becomes inoperable due to loss of both circulators. Similarly, Section 4.0 and LCO 4.3.1 of the current FSV Technical Specifications appear to imply the same 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> limitation for loss of both steam generator sec-tions in one loop. TSUP draft LCO 3.5.1.1 ACTION a allows 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to restore an OPERATING but inoperable loop to OPERABLE status before imposing the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> limit to effect an orderly shutdown. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowance needs to be justified since it differs from the original Technical Specifications. Since PSC currently recognizes the safety function as being accomplished only by the Pelton wheel drive, operatirg on feedwater (DBA 2) or firewater, the reference to "0PERATING loops" needs to be deleted since the Pelton wheel drive is not used for opera-tion at POWER. The second paragraph in the Basis for LCO 3.5.1.1 should also be deleted since the OPERABILITY or OPERATING status of the helium circulator on steam drive or on water turbine drive supplied by sources other than feedwater or firewater is not relevant to this LCO. PSC should also modify DEFINITION 1.23. OPERATING IN OPERATION, to read as follows (new wording underlinedj.

"A system, subsystem, train, component or device shall be OPERATING or IN OPERATION when it is OPERABLE r,er DEFINITION 1.22 and actually performing its specified safety function (s)."

OPERABLE is understood by the Staff to include the operabiliev of the full complement of equipment with no loss in redundancy.

• 6. The portion of the previous comment with regard to OPERABILITY versus OPERATING status also applies to LCO 3.5.1.1 ACTION b. A primary coolant loop can only be OPERABLE at POWER, LOW PO'.'ER and STARTUP since the steam drive has no safety function in safe shutdown cooling, and since the firewater safety function should not be OPERATING under any of these conditions.

• 7. LCO 3.5.1.1 ACTIONS c and d allow longer response times than allowed in Section 4.0 and LCOs 4.2.1 and 4.2.2 of the existing FSV Technical Specifications. LCO 3.5.1.1 ACTION e should be made clear as to whether OPERABLE refers to the Pelton wheel drive or to the bollar

12 ,

feed pump drive source. This ACTION needs to be more explicit.

i *8. In LCO 3.5.1.1 ACTIONS h and i, the word "immediately" should be replaced with the words "within 10 minutes."

• 9. In LCO 3.5.1.1 ACTION 1, the cited figure numbers should be 3 5.1-

1. 3.5.1 2 and 3.5.1-3 and not 3.4.4-1, etc.

• 10. In SR 4.5.1.1.a.2.a), a surveillance for the use of the turbine water removal tank overflow should be included consistent with the last paragraph on the second page of the Basis for LCO 3.5.1.1.

• 11. In SR 4.5.1.1.a.2.b), bearing water makeup pump P 2108 was omitted and needs to be included.

• 12. SR 4.5.1.1.a.3 should include the exact same provisions as the NRC redraft of SR 4.5.1.1.a.2.c as given in Enclosure 2 to the NRC NRR letter dated December 27, 1985, or PSC should provide justification for not including.

• 13. bR 4.5.1.1.b.2.a) should be replaced to read as follows:

"The main steam ring header to main steam pipins veld for one steam generator module in each loop, and."

The proposed readings for SR 4.5.1.1.b.2.a and b should be renumbered as b and c respectively. These changes are consistent with the NRC redraft version as given in Enclosure 2 to the NRC NRR letter to PSC dated December 27, 1985.

• 14 In SR 4.5.1.1.b.3, the word "in" in the first line should be changed to "is".

• 15. In SR 4.5.1.1.b.3.b), the words "and at least the preliminary" in lines six and seven should be deleted and replaced by the word "the" only.

• 16. In the first paragraph an the first page of the Basis for LCO 3.5.1.1, the words " condensate or" and " condensate and" should be deleted since PSC contends that safe shutdown cooling capability is not dependent on the availability of condensate.

17. The third paragraph on the first page of the basis for LCO 3.5.1.1 should also indicate that the apparent need for "two circulators,  !

operating with emergency water drive, supplied with feedwater via the I emergency water header" also applies to the maximum credible depressuri. l

, :ation rate is implied by FSAR Section 14.4.3.2. PSC should address

" protection against single failures" for such credible events. ,

l

• 18. The fourth paragraph on the first page of the Basis for LCO 3.5.1.1 uses mixed terminology compared to the rest of the Basis. The words " main steam / water section" should to replaced with " economizer-evaporator superheater section."

13

19. The third paragraph on page two of the Basis for LCO 3.5.1.1 should clearly indicate that the boiler feed (feedwater) pumps are not quali-fled as SAFE SHUTDOWN COOLING equipment. The probability value "E-7" should be rewritten as "1.0E 7" and justified. A probability estimate should also be supplied for the maximum credible depressurization event.

• 20. In the first paragraph on the third page of the Basis for LCO 3.5.1.1, delete the first two sentences and the word "also" in the third sentence. The backup bearing water supply system is not currently covered in the Technical Specifications and should not be credited in the Basis.

• 21. In the third paragraph on the third page of the Basis for LCO 3.5.1.1, the words "at 31 days and REFUELING CYCLE intervals" should be changed to read "at 31 days, 92 days and REFUELING CYCLE intervals."

• 22. On page four of the Basis for LCO 3.5.1.1, the equipment redun-dancy criteria is tied to the CALCULATED BULK CORE TEMPERATURE exceeding a value of 7600F. Attachment 4 to P 86169 discusses this parameter.

This discussion fails to address specifically how the operator is sup-posed to calculate and use this parameter for assessing the Technical Specifications applicability on line as the plant operates or changes operating conditions. PSC has not, for example:

(1) Presented the verification and validation methods and documentation to support estimating the CALCULATED BULK CORE TEMPERATURE (including the use of historical data);

(2) Identified the applicable provisions of Specification 6.8 for the Admints:rative Controls of this calculation and the supporting calculational procedures; and (3) Provided justification for not providing Technical Specifications l

Administrative Controls for this calculation.

PSC's responsa to this comment should be coasistent with the NRC gui-dance on the revision to the existing Specification 4.1.9 as provided in the NRC letter, K. L.11ettner to R. O. Williams Jr. , dated December 5,

! 1986.

) 23. FSAR Appendix C.15 cites the water turbino automstic start as an j " additional" engineered safety feature protection system, but this i

' feature is not cited elsewhere in the FSAR nor outside the basis in this Specification. PSC should either delete the reference to this system in the FSAR and basis or provide more information and appropriate specifi-a cations if any credit is to be taken for it in the safety analysis, 1

including the demonstration of compliance with the design criteria as is currently bein6 done.

24. Son comment 1 to LCO 3.7.1.1 with regard to etarifying the Basis on deptessurized cooling requirments.

l l 1 i

l

l 14 l 1

PROPOSED RESOLUTIONS j 1. PSC needs to incorporate the required change.

1 2. PSC needs to consider the appropriatoness of a separate LCO for DBA-2 cooling requirements and of a separate category for DBA 2 cooling equipment. At a minimum, PSC~needs to provide necessary clarifications l in the Basis to indicate that the drive sources (that is, one or more of l

1 the non Class I boiler feed pumps supplied in turn by either the non-Class I condensate pump [s] or the non Class I auxiliary boiler feed

pump [s]) are not qualified as SAFE SHUTDOWN C00LINC components.

i 4

3. PSC needs to incorporate the required change.

4 PSC needs to indicate concurrence with the NRC interpretation. PSC also needs to rewrite their response to Action 27b.

", 5/6. PSC needs to incorporate necessary changes and provide requested justification.

. , 7. PSC needs to provide justification of allowed outage time, and PSC i

needs to provide clarification and changes to the LCO to reflect clarif-ications with regard to OPERABLE components.

8. through 16. PSC needs to incorporate the required changes or to pro-j vide justification for not doing so if appropriate.

! 17. PSC needs to provide necessary clarifications.

18. PSC needs to incorporate the required changes.

1 i

19.

PSC needs to provide clarifications and necessary changes.

{ 20/21. PSC needs to incorporate the required changes, i

22. PSC needs to provide the required information and to assure con-4 sistency with the guidance in the cited NRC letter dated December 5, i 1986.

j 23.

PSC needs to provide clarifications and any necessary changes to i

the FSAR.

24 See the proposed resolution to comment i to Lc0 3.7.1.1.

i NRC COHMENTS LCO 3.5.1,2

• 1. In LCO 3.5.1.2.a.2, the words " including two CPERABLE flow paths" j should be inserted after the last word "0PERABLE" in the condition statement. This comment is the same as NRC Comment I to LCO 3.5.1.1.

• 2. In LCO 3.5.1.2.b.1 and .2 the lead word "One" should be changed to 1,

"A".

1 4

J f

. . .= . ~

1 15 4 3. In LCO 3.5.1.2.b.1, the drive source is not qualified for safe shut-

down cooling. See Comment 2 to LCO 3.5.1.1.

t

• 4. In LCO 3.5.1.2 ACTION a, the word "immediately" should be replaced j

with the words "within 10 minutes", and the words " control rod move-4 ments" should be replaced with the words "any evolution."

s *S. In LCO 3.5.1.2 ACTION b, the words " control rod movements" should j

be replaced with "any evolutions." Also, the word "immediately" should be replaced with tha words "within 10 minutes."

)1 PROPOSED RESOLUTIONS 1./2./4./5. PSC needs to incorporate the required changes or propose 1

j justification for not incorporating.

). 3. PSC needs to consider the appropriateness of a separate LCO for

}

DBA-2 cooling requirements and of a separate category for DBA-2 cooling i

equipment. At a minimum, PSC naads to provide necessary clarifications i

in the Basis to indicate that the drive sources (that is, one or more of i the non Class I boiler feed pumps supplied in turn by either the non-Class I condensate pump [s] or the non Class I auxiliary boiler feed l

pump [s)) are not qualified as SAFE SHUTDOWN COOLING components.

l NRC COMMENTS - LCO 3.6.2.1

, *1. LCO 3.6.2.1 ACTION a. and the first paragraph of the Basis for LCO i 3.6.2 allude to the 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of allowable operation with one LCS loop OPERATING: however, FSAR Section 5.9.2.4, page 5.913, does not provide i

specific details on the analysis that justifies the allowance of the 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> grace period.

• 2.

In LCO 3.6.2.1 ACTION a, the words " control rod movements resulting in" should be deleted.

i

• 3. In SR 4.6.2.1, PSC has failed to include the provision for scanner j

alarm functional tests and calibrations as provided in the NRC redraft in Enclosure 2 to the NRC NRR letter dated December 27, 1985.

I

! *4

! In the Basis for LCO 3.6.2, the last sentence of the second para-graph on the first page does not indicate how many tube failures were considered, nor does the following discussion therein provide concrete temperature limits'as a function of the number of failed tubes.

• 5

!. On page two, second paragraph, of the Basis for LCO 3.6.2, the ISI/IST program needs to be clarified as to how it " verifies OPERABIL.

ITY" of the subject barriers between safety and non safety portions of LCS.

Also, clarification is needed as to how tube flow rate instrument

accuracy is assessed without imposing Surveillance Requirements.

I 4

i i

! i i

s

- 16 PROPOSED RESOLUTIONS

1. PSC needs to update the FSAR to support the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period for shut-down.

2. PSC needs to incorporate the required changes.

3. PSC needs to include the recommended surveillances or provide jus-tification for not doing so.

4./5. PSC needs to provido necessary clarification.

NRC COMMENTS - LCO 3.6.2.2

• 1. In LCO 3.6.2.2 APPLICABILITY, the pound sign (*) footnote should be further explained as to why operation should be allowed in a degraded mode. This appears to be a new licensing condition which has not been

! supported by adequate justification.

• 2. In LCO 3.6.2.2 ACTION, the words " control rod movements" should be replaced with the words "any evolution."

t

• 3. On page two of the Basis for LCO 3.6.2, the fourth paragraph should be evaluated and revised based on comment 1 above.

PROPOSED RESOLUTIONS

1. PSC needs to provide justification for operation in a degraded condi.

tion.

2./3. PSC needs to incorporate the required changes.

NRC COMMENTS LCO 3.6.3

• l. In LCO 3.6.3 ACTION, PSC should delete subheading "a." and delete the words " control rod movements."

• 2. In SR 4.6.3, PSC should change SR 4.6.3 to cover functional tests and instrument surveillances ustng the wording given in SR 4.6.3.c in the NRC redraft in Enclosure 2 to NRC NRR letter dated December 27, 1985.

• 3. The FSAR should be updated with a discussion of the temperature limits associated with the LCS to correct the deficiency noted in the Basis for LCO 3.6.3.

i PROPOSED RES01.UTIONS 1./2./3. PSC needs to incorporate the required changes or provide jus-i tification for not doing so.

, , 17 NRC COMMENTS - Section 3/4.7

1. The title for TSUP draft Section 3/4.7 should be reduced simply to PLANT SYSTEMS. SAFE SHUTDOWN COOLING SUPPORT SYSTEMS can not be support systems if in fact the designated system has to oeerate in order for the

" emergency core cooling" function to be performed. Similarly, the sys-tem should not be so defined if in fact it does no: support " safe shut-down cooling." The current title is a misnomer. As also indicated in other NRC comments, certain Specifications (LCOs 3.7.5 and 3.7.8) fail to meet the intent of designation as PLANT SYSTEMS. and these Specifica-tions should be relocated to other appropriate TSU? draf t Sections.

2. FSAR Sections 2.5.1 and 10.3.9 (page 10.3 8) define the South Platte River, Fort St. Vrain Creek, and the system of six shallow wells as con-stituting that which appears to be an Utimate Heat Sink. Per the licensing basis, the onsite storage ponds capacity will be exhausted after about 11 days thus requiring use of offsite sources to replenish the storage ponds. The offsite sources, or ultimate heat sinks, are not currently addressed in TSUP draft Specification 3/a.5.4, and t'he Basis for this Specification does not cite the 11 day limit on onsite capacity cited in the FSAR. Why is there no Utimate Heat Sink Specification con-sistent with the licensing basis and similar to that exemplified by W.

STS Specification 3/4.7.57 b

PROPOSED RESOLUTIONS

1. PSC needs to revise the TSUP draf t as l'ndicated.

2. PSC needs to ensure that the TSUP is complete and consistent with the licensing basis.

NRC COMMENT LCO 3.7.1.1

1. On page 2, Section D, Attachment 1 to P 86169. PSC states that "with the singular exception of the Rapid Depressurization (DBA 2), one circu-lator on pelton wheel drive assures safe shutdown cooling." The FSAR apparently addresses two depressurization events which require the operation of two circulators on feedwater drive of the Pelton wheels.

One depressurization event is postulated to occur instantaneously, i.e. ,

the DBA 2 as described in FSAR Section 4.11, and the other is assumed to occur i.e.,

with an exponential time constant of 1600 seconds (or 26 minutes),

the largest credible PCRV leak rate as described in FSAR Sections 14.4.3, 14.7, and 14.8.

FSAR Section 14.4.3.2 ses:es that the maximum credible depressurization accident, which would bring the PCRV to atmos-pheric pressure in about 80 minutes, is adequately cooled by two circu-  ;

lators on water turbine drive and that the DBA 2 is analyzed to provide I a bounding case. The FSAR is unclear as to what depressurization rate constitutes the crossover point between needing two circulators versus only one circu~.ator operating on feedwater turbina drive. The FSAR is equally unclear as to what depressurization rate, if any, is accommo-dated by one or more circulators operating on either condensate or fire-water. Has this analysis been performed? If so, why is it not

18 presented in the FSAR? The minimum depressurization rate requiring the operation of two circulators on feedvater-turbine drive establishes the PCRV leak size which in turn establishes the basis for estimating the event occurrence probability and thereby the need for circulator redun-dancies. The GA probabilistic analysis for the DBA-2 occurrance fre-quency appears to be based on a possibly optimistic use of the rupture frequency for PWR vessels. PSC should reexamine the fundamental logic behind both the WASH-1400 and British PWR vessel rupture studies. The assumptions behind this logic may not be appropriate for applying the PWR vessel rupture frequency to FSV.

2. This LCO/SR should be moved to TSUP draft Section 3/4.5 since the components covered must operate for emergency core cooling during DBA 2.

These components are essential, not merely supportive, to the operation of SAFE SHUTDOWN COOLING drives during DBA-2, but the components are not qualified to be on the SAFE SHUTDOWN COOLING equipment list. Therefore. -

the second sentence of the first paragraph on page 3/4.7-3 of the Basis should be deleted because it implies incorrectly that the boiler feed pumps effect SAFE SHUTDOWN COOLING and states incorrectly that the con-densate pumps can do the same. It appears that the boiler feed pumps are used to effect " emergency core cooling" during DBA-2 per STS termi-nology and that the condensate pumps are used to effect " residual heat removal" per STS terminology. However, for the boiler feed pumps to be operable, either the non-Class I condensate pump (s) or the non Class I auxiliary boiler feed pump (s) must be operable to supply an adequate net positive suction head for the boiler feed pumps. Therefore, additional operability and surveillance requirements appear to be needed to assure the availability of the DBA-2 cooling system.

3.

Should REFUELING be added to the APPLICABILITY statement since PSC has no where indicated the value of the CALCULATED BULK CORE TEMPERATURE which must exist in order to enter the REFUELING mode?

4. FSAR Section 6.3 and FSAR Appendix Sections C 41 and C.44 need to be revised to be consistent with FSAR Sections 14.4.3.2 and 14.11.2.2 with regard to the number of operable circulators on feedwater drive needed for depressurized core cooling. FSAR Appendix Section C.44 cites an incorrect figure number in FSAR Section 14.4 Apparently, the cited figure is no longer included in the FSAR and has been replaced by Figure 14.11 11.

5. Previously recommended changes to this Specification per NRC letter, K. L. Heitner to R. F. Walker, dated May 30, 1986, remain in effect.

PROPOSED RESOLUTIONS

1. PSC needs to provide the necessary clarifications and changes to the Basis for LCO 3.5.1.1 and LCO 3.7.1.1 and the FSAR with regard to depressurized cooling requirements arising from depressurizations other than DBA 2. PSC needs to clarify and justify the use of PWR vessel rup-ture frequencies for assessing the frequency of FSV accidential depres-surizations. PSC needs to justify not applying ASME CodesSection XI inservice inspection requirements to FSV penetration closures.

. , 19

2. PSC needs to relocate the LCO/SR and to revise the Basis as indi-cated. PSC needs to expand or add necessary LCOs/SRs to cover the DBA 2 cooling system requirements in an effective manner.

3. PSC needs to provide clarification on the value of the CALCULATED CORE BULK TEMPERATURE with respect to REFUELING.

4 PSC needs to revise the FSAR as indicated.

5. PSC needs also to comply with previous comments on the LCO/SR.

NRC COMMENT - LCO 3.7.1.2 Does a steam / water dump actuation always occur prior to all the possible sequences leading to the need for emergency core cooling, including SAFE SHUTDOWN COOLING? How does the actuation mechanism for.the feedwater isolation valves differ between the steam / water dump and all the possi-ble sequences involved in requiring emergency core cooling? Does the latter include manuel and remote-manual operation, possibly via dif-ferent redundant electrical circuits in remote manual operation? Do the proposed LCO and SR accurately depict and verify OPERABILITY under all the conditions in which the feedwater isolation valves are required to operate for emergency core cooling? If the answer to the last question is "No," PSC needs to propose an applicable and apptopriate Specifica-tion for the feedwater isolation values to be included in TSUP draft Section 3/4 5. Even if "Yes", if these valves are absolutelv essential to the operation of the emergency core cooling system, an appropriate .

Specification should still be included in the proper TSUP Section 3/4.5.

Finally, FSAR Section 10.2 is very unclear as to when or whether the emergency feedwater header has to be in use during STARTUP, SHUTDOWN, and REFUELING. Use of the emergency feedwater header would apparently imply that the feedwr.ter isolation valves are closed. Since emergency core cooling could apparently be required from STARTUP or SHUTDOWN, and perhaps even from REFUELING for which the required value of the CALCU-LATED CORE BULK TEMPERATURE is not defined, the APPLICABILITY statement for the feedwater isolation valve Specification should be expanded beyond POWER and LOW POWER to situations in which the isolation valves may not already be closed.

Previous NRC comments transmitted by the NRC letter dated May 30, 1986, continue to apply.

PROPOSED RESOLUTION PSC needs to provide the requested clarification and any TSUP draft revisions as necessary and appropriate.

NRC COMMENT LCO 3.7.1_.5

• In LCO 3.7.1.5.b, the reference to Specification 3.0.6 should be changed to 3.0.4. In SR 4.7.1.5, the reference to Specification 4.0.6 should be changed to 4.0.4 These changes are consistent with the

, ~

s -

. 20 difference between PSC's April 1985 draft and the.recent NRC markup of the November 1985 final draft.

PROPOSED RESOLUTION 4

PSC needs to incorporate the required changes.

NRC COMMENT - LCO 3. 7.1.6

• In LCO 3.7.1.6 ACTION, change the words " control rod movements" to "any evolutions." .

PROPOSED RESOLUTION PSC needs to incorporate the required change.

NRC COMMENT - LCO 3.7.3

.If the instrument air system is absolutely required for the functions defined in the Basis, especially the operation of the pneumatic valves required for SAFE SHUTDOWN COOLING, then this Specification should be relocated into TSUP draft Section 3/4.5 as an essential part of the emergency core cooling system. Previous NRC comments transmitted by NRC letter dated May 30, 1986, continue to' apply.

PROPOSED RESOLUTION PSC needs to relocate the LCO/SR as deemed appropriate and necessary for the system function.

NRC COMMENT - LCO 3.7.5 The TSUP draft Specification for the Primary Coolant Depressurization system is functionally equivalent to the STS Specifications for Contain.

ment Depressurization and Cooling Systems. The Fort St. Vrain Primary Coolant Depressurization has no PLANT SYSTEMS function, only containment equivalent functions. Previous NRC comments transmitted by the NRC letter dated May 30, 1986, continue to apply.

PROPOSED RESOLUTION The LCO/SR needs to be relocated by PSC to TSUP draft Section 3/4.6.

NRC COMMENTS . LCO 3.7.8

1. The ACM diesel generator specification should be relocated to TSUP draft Section 3/4.8. The ACM is not a PLANT SYSTEM, specifically not a fire protection system.

2. The Basis for the proposed specification 3/4.7.8 is not consistent with that given for the existing specifications 4.2.17 and 5.2.20. The Basis for the proposed Specification 3/4.7.8 must be revised to elim.

inate the implication that the intended function of the ACM is to effect

. ~

21 SAFE SHUTDOWN COOLING. Instead, the specific functions of the ACM as given in FSAR Section 8.2.8.2 need to be included in the Basis, and the consequence goal of the ACM needs to specified per FSAR Sections 8.2.8, namely, insuring that "the conditions of public health and safety conse-quences, analyzed and presented in Design Basis Accident Number 1 in (FSAR) Appendix D.1, are not exceeded in the case of ... disruptive faults or events in congested cable areas." The fact that the risk goal for the ACM is an " acceptable substitution" for the cold shutdown (or safe shutdown) goal per 10CFR50, Appendix R, should also be noted, but no statement should be made either indicating or implying that the NRC's agreement to an " acceptable substitution" represents " equivalency" to the requirements of 10CFR50, Appendix R. The ACM provides for " core suberiticality and safe contair. ment" which is not equivalent to a condi-tion of " safe shutdown" without fuel damage.

3. FSAR Appendix C and FSAR Section 1.2.2.9 need to be revised to present a protrayal of the ACM functions and consequence goal consistent with that presented in FSAR Sections 8.2.8 and 14.10 and FSAR Appendix D.

Use of such terms as " safe shutdown," " safe shutdown cooling,"

" emergency cooling," and " safe condition" to refer directly or indirectly to the ACM function or to the associated operation of PCRV LCS should be eliminated in the FSAR Appendix C and FSAR Section 1.2.2.9. In addition, FSAR Appendix C.48 needs to be revised to remove the implied capability of the ACM diesel generator to perform "the sequential programming of essential electrical loads."

4 Previous NRC comments transmitted by the NRC letter dated May 30, 1986, continue to apply.

PROPOSED RESOLUTION

1. through 4 PSC needs to revise and relocate the Specification and to revise the FSAR as indicated.

NRC COMMENT - LCO 3.9.1 The requirecent that the CORE AVERAGE INLET TEMPERATURE must be less than or equal to 1650F is not sufficient to define safe conditions for REFUELING.

The relationship between the allowed entry into the REFUEL-ING mode and the value of the CALCULATED BULK CORE TEMPERATURE needs to be specified. The cooling capacity during REFUELING should be specified in terms of the preferred equivalent " residual heat removal" system con-figuration consistent with the approach used in the W-STS Specification 3/4.9.8 and with the implied assumption that such capacity is available per FSAR Section 9.1.1.4 The fact that PSC failed to identify within the FSAR the preferred equipment configuration necessary to meet the minimum operating conditions for REFUELING should not be construed as relieving PSC from the responsibility of specifying the needed equipment OPERATING and OPERABILITY conditions within the Specifications. If PSC prefers to divide up or realign Specifications within TSUP draft 3/4.9 to respond to these comments, such revisions are acceptable provided

~ .--

. . 22 that the above comments and those comments previously provided by NRC in the NRC letter dated May 30, 1986, are incorporated.

Currently, core conditions during REFUELING are unclear with regard to expected heat loads and cooling capacity. The re fore , FSAR Section 9.1 should be revised to include figures and perhaps tables that provide a correlation of typical expected values for the CALCULATED BULK CORE TEM-PERATURE, the calculated decay heat fraction normalized to rated reactor power, and the CORS AVERAGE INLET TEMPERATURE as a function of time after shutdown. Such correlated data would be useful to demonstrate expected core heat loads and cooling capacities in SHUTDOWN preceding

• entry into REFUELING. Such information should be available for SHUTDOWN both immediately from full power at the end of an equilibrium cycle and following an " orderly" shutdown from full power at the end of an equili-brium cycle. This comparative information should clarify the conditions existing within the reactor and core following entry into SHUTDOWN and before REFUELING.

PROPOSED RESOLUTION PSC needs to revise the Specification and FSAR as indicated.

NRC COMMENT - SR 4.5.1.1 In addition to our previous comments, we request that you consider further revisions for Specification SR 4.5.1.1. These revisions are necessitated by concerns raised in Reportable Occurrence 50-267/

86-026 with regard to demonstrating the adequate supply of firewater simultaneously to both the belium circulator Pelton wheel and the steam generator economizer-evaporator-superheater (EES) section. SR 4.5.1.1.b.1 needs to be modified to be performed in conjunction with SR 4.5.1.1.a.4.b).

In SR 4.5.1.1.b.1, " proper flow" through the emergency -feedwater header and the emergency condensate header also should be quantified in terms of the accident conditions. This means the surveillance test conditions should be at the required flow of simulated firewater with accident condition back pressure in the EES section and at circulator operation 3% (or 3.8%)

rated helium flow on throttled condensate to simulate firewater pump dis-charge. In effect, SR 4.5.1.1.2.4.b) and SR 4.5.1.1.b.1 should reproduce test conditions equivalent to Test T-30 and/or T-30A so as to verify the continued efficacy of the firewater cooldown on a periodic basis. Test conditions such as throttled condensate pressure and EES back-pressure need to be stipulated in the Basis for SR 4.5.1.1.

PROPOSED RESOLUTION PSC needs to revise the specification as suggested.

. Amendmsnt No. Enclosura 2 4 -

Page 3/4.5-SAFE SHUT 00WN COOLING SYSTEMS bb/Y*"I'

• 3/4.5.1 SAFE SHUTDOWN COOLING EQUIPMENT i

LIMITING CONDITION FOR OPERATION ,

3.5.1.1. a. Two primary coolant loops shall be OPERA 8LE, each with at least:

1. One helium circulator OPERABLE, and

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

OPERA 8L .

lb *

1. **D *

b. For OPERABLE helium circulators, the following safe O PE R A BLE shutdown cooling drives and auxiliary equipment shall be OPERABLE:

-flow pat hs

1) A safe shutdown cooling drive with the capability of providing the equivalent of 8000 rpm circulator speed at atmospheric . pressure - to two circulators 7 :g, simultaneously, 1 4s A E#\C gV f 2) JW6 safe shutdown cooling drive with the capability ggM of providing 3% rated helium flow at operating

g. pressure with firewater supply, including two O Qs ~'(y i OPERA 8LE emergency uter booster pumps and two OeeRABLE fiow neths.

~ 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 r - - -r-- ,. - - - - -- - ,, c --r - , , - , , -

Amsndmant No.

page 3/4.5-DRAFT FE8 ?s7993  ;

6) The supply and discharge valve interlocks ensuring tutomatic water turbine start capability.#

APPLICABILITY: POWER, LOW POWER, STARTUP* and SHUTDOWN

• ACTION:

. a. With only one OPERABLE primary coolant loop. -d un k-t?

? ::;; 0FCFAT! 0, , ,,;.;. L . .,v,,u . 6 :..y  ;. OP ".".0'.8 at: .; t h i ,, 72 ....r; the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Q be in at least SHUTOOWN within f it only r to e e OP BL a p m y co lan loop,

-ope ting op t E Gs tu n ho , r b i a lea HUT wt n e ne 4 hou . R tore o1 st L. at w n ho s t nit lo of a E 1 or i at 1 ast UTO wi i t xt 4 h urs, b[ With no OPERABLE safe shutdown cooling circulator drive capable of 8000 rpm equivalent, ::: ;r: th: ' nd dr' d 1: 0"EP*?tE :t:'u: -!iS'- "- aart 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />s-r: rE be in at least SHUT 00WN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c. With only one OPERABLE firewater-supplied helium cir~c ulator drive, . ;;t:r: :t '

st:tu

.. m -- J.:.:. t: 0"E"*"' $

!th'- 72 h... ; :". be in at least SHUT 00WN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

M[ 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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or be in at least SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

With less than the above required OPERA 8LE bearirs water accumulators, restore the inoperable equipment to (PERA8LE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or be in at least SHUTOOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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.

j* .

Amendment No.

Page 3/4.5-DRO.FT With FEB 2 819es ,

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

With loss of both redundant OPERABLE components or flow paths required by LCO 3.5.1.1.b.2, 3, or 4, (e.g., with )

- t both turbine water removal pumps or with both emergency water booster pumps inoperable), be in SHUT 00WN

' -- - '" "

• 9 itt: . to mi n vT e]

[ With no OPERA 8LE or OPERATING primary coolant loops, be at least SHUTOOWN 4-- Mf:t:?7 nd restore at least one loop to OPERATING status within 90 minutes, or depressurize the PCRV in accordance with the applicable wlrk requirement below.

If forced circulation is restored 1o within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of initial loss, depressurization may be discontinued.

j

• ') 1. As a function of reactor THERMAL POWER equal to or greater than 25% prior to SHUT 00WN, as delineated in Figure 3. -1.

2.

As a function of CORE AVERAGE OUTLET. TEMPERATURE for reactor thermal power less than 2 prior to SHUTDOWN,asdelineatedinFigure3.f.[5" 2.

3.

r8 As a function of time from reactor SHUTDOWN as delineated in Figure 3. . -3. .

fI 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 N V e.e

• f y i n 3 Iksi sw O PE R A BLE

-f lo w poTh axisTs b e~1 w e e. n 'C h e tv k: o e. w 2T er emeuul T a w (4 l o v e i- -f t . w 3 J T h c_ reaeT ..-

b v 's 12

• g g sump,2nd

. l Amendment No.

Page 3/4.5-lid ltiTy nJ P 11 E 0f) FES 2 6 %

Performing a start test of bearing water makeup

- pum;,f 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 10 will be tested during even during odd numbered cycles and circulators 1A and 1C numbered cycles, by demonstrating operation on water turbine drive by:

, a) Verifying an equivalent 8000 rpm (at atmospheric pressure) feedwater on two circulators simultanecusly or.

motive power using the emergency feedwater header, arid 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 t that will over 4 inspection periods, each circulator ,

be testedaccessible components, once. Other helium circulator {

disassembly without further than required to inspect these wheels, shall be visually examined, and

1

' i Ar:endment No. '

Page 3/4.5-DRf ~

w

{3 b) At least fee ^ . -

nT 10*.' of primary coolant pressure L

w a boundary which bolting and other structural bolting has been removed for the inspection above M 7 and which is exposed to the primary coolant shall be nondestructively r t. tested for

. . , identification of inherent or developed defects.

0 lLG c) Reports w

0 2 Within 90 days of examination completion, a N (

[ Special Report shall be submitted to the NRC in accordance with Specification 6.9.2.

This W I report shall include the results of the helium i " circulator examinations, n5" b.

The steam generators shall be demonstrated OPERA 8LE:

v

-r y 1. At n g least once per 18 months be verifying proper flow r through the emergency feedwater header and emergency condensate header to the steam generator sections.,

I L j J * # 2.

At'least once per 5 years by volumetrically examining

• YD V

the accessible portions of the following bimetallic y3 3 ~4 M for indications of subsurface defects:

welds

,-w p r3,

9) The . main steam ring header collector to collector drain piping weld for one steam generator module in each loop, and c.,$[ The f..f{

v a. v n same two re examined at each interval.

steam generator modules shall be

< J-i F #- .f The initial examination shall

• SHUTUCWN cycle 5.

or REFUELING prior to the beg.fuel inning ofb

the bimetallic This initial examination shall also include welds described above for two additional steam generator modules in each loop.

.[

3. Tube Leak Examination e .

l .

Each time a steam generator tube i

• leak, specimens from the accessible plugged due to a connected subheader tubes to the leaking inaccessible tubes shall be metallographically examined.

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

1 i

l l  :

i

, 1

Amendment No.

Page 3/4.5-OP. ' u r .

FEE .0 e :ce:

A study shall be performed to evaluate the siz.e 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 An engineering evaluation shall be performed to determine the acceptability of:

1) Any subsurface defects identified in Speci fication 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 Soecification 6.9.2. This report shall include the estimated size a-and elevation of the leak (s), :-d :t ' ;;;tt t k e + ha 14 *-- yo results of the metallographic and engineering analyses perfo rmed, the postulated cause of the leak if identified and corrective action to be taken.

I 4

l

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

b k~ . . . .

9.# I .

FEB.: N::

5 I I I I I I i E

8 f TBE AVAluSLE PRIOR 70 INfflAT198 07 PCRV

= ountuunuArios iness 70nc 0 cincutATion o 4 IR Leff FRSW A POWERES CestfflGR AT PSV

,h 70 SE UgES.FOR

~

PO.WE A LEYtta ESUAL TO OR __

. ma r. . .

3 I

g 3 o I i

E  %

W  % 1 g2 l e .. I 5 .

2 3 .

\

t t . \

5  !'

i E

>=

d l

I 0 _

l 20 30 44 50 50 70 m 90 100 REACTOR THERWAL POWEA - %

Tane Ash Prter to faitiation of PCRV Depmsezation When Forced Circulation in Lost frein a Powered Condition at FSY Figure 3.5.1-1 1

l 1

l l

,n, . . 1 .~ ~ .

F E B 2 ; c :.;

9 i

I I I I I I 2

=

I I I I E3 i Test AVAILASLI PRIOR 70 infftAft04 0F -

j k PCRV 04PREEM8Al2ATM A8 A FUNCTION

- er AvenAss come ovfLETTserEnAfons AT TNS ONOET SF A LOPC 2 y

\ TO N USEB FOR PWWtn LEVELS LES TNAN ISE i

5 h i

l i

3 \

0 o  %

A

{5 h' .

o

~

NN ~~

e f w -

4

- \- ,

5 5  %

<3 g \+ ,

N gi A

400 500 63 7N 800 SW 1000 1100 IM 13 3 H00 1500 AVEMAGE CORE OUTLET TEMPERATURE 'F of Amage Con Outlet Tempentun at the Orwt i I

Figure 3.5.1-2 l

l

l l

Q Q .* ~-

FE8 2 E : -

g 80 , , , , , , , , , , , , ,

s E

4 70 TIME AVAILA8LE PRIOR 70 INITIATION OF /

g -- PCRV OEPRESSURIZATION WHEN FORCEO ~

Q CIRCULATION 13 LOST FROM A SHUT 00WN CONDITION

/

ts 60 /

E g TO BE USED FOR A SHUTOOWN CON 0lTION

> ONLY ii

\ i t

l s /

< ** /

) / -

3 -

- d p r 9 # -

E 3 20

[ ~ c E >

/

s to ' /

t 0

200 W 400 500 600 700 800 m 1000 TIME FROM REACTOR SHUTOOWN - HOURS Tune Available Prior to Initiation of PCRV Depressurintion When Forced Circulation is Lost from a Shur Down Condition F'gu~e 3.5.1 -3 i

Amendment No.

Page 3/4.5-SAFE SHUTDOWN COOLING SYSTEMS

~

3/4.5.1 SAFE SHUTDOWN COOLING EQUIPMENT g g :-

FtB 2

LIMITING CONDITION FOR OPERATION 3.5.1.2 a. At leastat including one primary coolant loop shall be OPERABLE, least:

1. One helium circulator OPERABLE, and 2.

One steam generator section (either the economizer-c l u el ,* ., $

evaporator superheater (EES) or reheater) OPERABLg 7wo OPER AB ta b. For at least one OPERABLE helium circulator, the following flow paThr emergencyg drives and auxiliary equipment shall be OPERABLE:

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

2. sa fe providing 3% shutdown cooling drive with the capability of rated helium with flow at operating pressure firewater supply, including one OPERA 8LE emergency water booster pump and one OPERABLE flow path,

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

• 4 The no,mai eearieg water s, stem. incieding one source of Ip,[g v bearing (P2105 orwater makeup P2108), and and one bearing water makeup pump 5.

O ?Q t ' The bearing water accumulator (T-2112, T-2113, T-2114, g T-2115) for the OPERABLE circulator (s). .

APPLICABILITY: STARTUP*, SHUT 00WN", and REFUELING Whenever CALCULATED 760 degrees F. BULK CORE TEMPERATURE is less than or equal to

Amendment No.

Page 3/4.5-C$d?'

ACTION: a. With less than the FEB2Atr-with forced circulation maintained, be in atabove required OPERA w t ru % to o ca'- '7,'and least SHUT 00WN restore the required equipment to OPERABLE

" ' " #I ' ' status prior to reaching a CALCULATED BULK CORE TEMPERATURE

.o f 760 degrees F,orsuspendalljperationsinvolvingCCRE ALTERATIONS or "^-**^1 -ad

=^"- -i. resulting in positive

. reactivitf changes, or ovement of IRRADIATED FUEL.

i,yeviuTi.@

b. With less than the above requirea UPLRABLE equipment, and with no OPERATING primary coolant loops, be in at least SHUT 00WN * :d'P d f and restore at least one loop to W e E4 *

• 10 OPERATING' status prior to reaching TEMPERATURE of 760 degrees F, or a CALCULATED BULK CORE 1.

Suspend all operations involving CORE ALTERATIONS or r^-ted --d

-^"^-- if resulting in positive reactive changes, or movement of IRRADIATED FUEL, and Y* 2.

Initiate PCRV depressurization in accordance with the time specified in Figures 3.5.1-2 or 3.5.1-3, applicable. as SURVEILLANCE REQUIREMENTS

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

I 1

Amendment No.

Page 3/4.5 DQ' Y FEB 2 o - i BASIS FOR SPECIFICATIONS LCO 3.5.1.1/SR 4.5.1.1 AND LCO-3.5.1.2/SR 4.5.1.2 ~

l One primary coolant loop, consisting of or.e helium circulator

- and one steam generator section ensures SAFE SHUT 00WN 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 loo auxiliary equipment which provides services to one (a)p.r aOne circulator, a d--" * = operating with motive power from either condensate header, ^* boosted firewater supplied via the emergency supplied via the or (b) feedwater or boosted firewater emergency feedwater header, provides sufficient primary coolant circulation for the pressurized condition.

SAFE SHUTDOWN COOLING is discussed in FSAR Section 10.3.9, m d^^ritt single failure considerations in Section 10.3.10, and ind" boosted firewater cooldown transients in FSAR Sections 14.4.2.1 and 14.4.2.2.

.Two circulators, supplied with feedwater operating via the with emergency water drive, emergency feedwater header, g,; I Q r.

provide sufficient primary coolant circulation followingra postulated S Design Basis Depressurization Accident (08A-2). tac et A-1 -Mir is a highly incredible event and protection against single failures is not a feature of FSV. ., ,

The requirements for c.re J u le.

adequate means for removingOPERABLE heat from the primary reactor j"l'provide steam generator (s) *" an coolant system helium flow which cools theto the secondary reactor coolant system. T Thei.w reactor core enters the steam generator at high temperature and gives up its heat to the ( F3 A E reheat steam section and main ittg'"itt i~t'tr. <,e T;ea Each steam generator consists of six identical individual I 4 ' 9

steam generator modules operating in parallel. Each module consists of o r-superheater section.a reheater section and an economizer-evaporator-Any one section provides sufficient heat removal capability to ensure SAFE SHUTDOWN COOLING C.,....,,o.~,.. ~ s-

Amendment No.

page 3/4.5-DRAFT During FEB 2 8 GS5 e BULK CORE TEMPERATURE greater than 760 F, one steam T* degre generator section in etch ,* *7)

This allows for a single failure loop is required to be OPERA 8LE.

means for removing decay heat. and provides an adequate 3 'e .e o

'T *- V

• n, 'a During SHUTDOWN with CALCULATED BULK CORE TEMPERATU 4 33 8

than or not equal toand required 760 degrees F and REFUELING, redundancy V Pj is 3 economizer-evaporator-superheatereither the reheater section or the section of one steam 'T* W*

generator primary can de used for shutdown heat removal from the st $z coolant. -

1 q i, W vi Safe Shutdown Cooling Drives w

~*y S-v

u. {

The requirement to .a

• 4 q

3 simultaneously at 8000 rpm be able to drive two circulators at conditions equivalent to

• - u o -

w ji 3

],

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

' supplied via the emergency feedwater header, kT esting to e demonstrate capability at conditions equivalent to atmospheric

• pressure is based on calculated helium density, reactar t *4

\ pressure stat *mant and circulator inlet temperature. A W hour action

- nighly incredible event, with a probabilityis acceptable because 08A-per year, of ~ '" e T o f lessthan-@E-1e.c o cc o r r o w e 8.

The requirement for with firewater supply ensures Safea drive capable of 3% rated helium flow event of Shutdown Cooling in the a loss pressurized of normal circulator steam turbine drive, in g o ,. .cderly h "the condition. A M our action ty e m !! 5 * *is statement 5 g , ,,, ,, ) for loss of redundancy in the Safe Shutdownused consistently in this Spe Cooling Systems.

This is acceptable in that an equipment failure in this relatively brief interval is unlikely.

complete loss In the event of a imediate shutdown ofisSafe Shutdown required, Cooling capability, an to place stable condition with reduced decay heat loads.

the reactor in a Circulator Auxiliary Equipment One turbine water removal pump has sufficient capacity to remove the water from two circulator the water turbines. Also, turbine water removal building sump is available if the tank overflow to the reactor nomal lost. pump flow path is

e - - - -- ,_.

. *j Amendment No.

. Page 3/4.5-3 DENFT

~

"d:periet . to. .i.; FE8 2 e ses 2pply Of b ;Tii.s -0t6, te th..:t:- eye +a=

t;;; c i rc:!: *^"* 4" *2^ pr'r;r o-nrnviriae a ennHnn~n enn14 g he ,, . S

' : th: ; ^.;;; 4s Leuoy ..gri, of t,i..iq .00: 7 g ccr:t r '::i.;te, ^

i; pr;;ff:d C2 water requirem;ents are , .i aa. ' Makeup bearing

! # norma,lly obtained from the feedwater provided system.

as "a backup A separate bearing water makeup pump is water surge t,ank. The bearing to supply makeup water to the bearing water makeup pump nonnally the condensate storage tanks. takes suction from the deaera

mergency bearing water makaup If this pump is inoperative, an pump can supply water at a reduced capacity from the condensate storage tank to the bearing wa ter surge tank. In an extreme emergency, filtered firewater either thecan bearinbe provided to the bearing water surge tank by water makeup pump.g water makeup pump or the emergency bearing water accumulator capable of supplying seconds at design flow rate wa ter for 30 bearingE 3 water is available. This is adequate if no other source of bearing for shutdown of the affected circulators without damage to the bearings.

The bearing water system, accumulators and the bearing _including water makeun the bearing water 1 functionally a"=a= ara >ggs respectively, tested at 31 day $nd REFUELING CYCLE intervals, to ensure pro er operation. There is redundancy in the bearing water accumulators and a 24nohour action s ta tement restoration time is provided.

acceptable components. considering the low likelihood of failures of these This is Au w or r tur e ut ne tart 's vent i wat tur RIP ate rb' e u os rt c ro sw i n i ee nce The 'or e ed erl cu r u (, LE, e i . e utoma ure to t r te a te tur ar on fe in t i '

th on o om h e an s

y ev .

r to o me i acc bl .

. u ur a ion s a nt

( ), l g , , m w at.r t u r i, ; .s e a vT.

, n ,1 ; , . 1 J ,u .br o. . a FsAR n r-c r e d .' I e d Ia 5* I l

1

. ~ _ _ ._ _ __ . .

Amendment No.

Page 3/4.5-

. Depres suriza tion D $cN pg g g g.3 In the unlikely event that all forced circulation is lost for 90 minutes, function of start of depressurization is initiated as a prior power levels, with 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> from 1001 RATED

- THERMAL POWER being the most limiting case. Operators will continue attempts to restore forced circulation cooling until 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after sources and the loss of forced circulation. Multiple using circulators flowpaths to establish forced convection cooling 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 vont stack filters to atmosphere. Start of depressurization times from various reactor power conditions are delineated in Figures 3.5.1-1, 3.5.1-2, Section 9.4.3.3 and Appendixand 3.5.1-3 D. and are discussed in the FSAR 1

R_edundancv Criteria 2

The use of 760 degrees F CALCULATED BULK CORE TEMPERATURE as a

division between the APPLICABILITY of Specification 3.5.1.1 l

verses 3.5.I.2 is explained as follows:

In

the FSV HTGR, the limiting parameter of interest is a core i

inlet temperature greater than 760 degrees F. The CALCULATED SULA CORE TEMPERATURE is maximum potential temperaturea conservative calculation of the 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 and 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 for failure consideration, accident prevention and mitigation without single cooling requirements. on the basis of the limited core I

i l

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T

.. . l Amendment No.

Page 3/4.5-bblA .N All forced circulation may be na 2 a n .

purposes provided that the time calculated for CALCULATED BU CORE TEMPERATURE to reach 760 degrees F is not exceeded.

However, if forced circulation is temporarily restored, a recalculation shall be perfortned, based on present conditions, to establish a new time period for CALCULATED BULK CORE TEMPERATURE to reach 760 degrees F.

also be taken out of service for maintenanceRedundant systems may testing provided or surveillance time to reach CALCULATED SULK degrees F may be recalculated as often as required.

equal CORE The to 760 TEMPERAT Steam Generators The steam generator reheater or EES sections can receive water from eitherfeedwa emergency the associated ter emergency condensate header the or per this Specification. header which are required to be OPERA 8LE by verifying flow System flow OPERABILITY is determined headers through each steam generator.from each of the aforementioned Bimetallic Weld Examination The steam generator incoloy 800 examination. and 21/4crossover Cr-1 Motube bimetallic welds between materials 'are not accessible for The bimetallic welds between steam generator ring header collector, the main steam piping, and the collector drain piping are accessible, materials, and operate at conditions not involve the same significantly different collector from drain theweld piping crossover tube bimetallic welds. The is also geometrically the crossover tube weld. Although similar to expected defects to occur, this specification allows for detectionminimal which of degradation affect bimetallic might result from conditions that can uniquely welds made between these Additional collector welds are inspected materials.

at the initial examination to establish should defects be a baseline which could be used, examinations subsequently be required.found in later inspections and add

s ,

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Amendment No.

Page 3/4.5 Tube I.eak Examination DMUT' F8 2 6 '033 During the lifetime of the plant, a certain number of steam generator tube leaks are expected to occur, and the steam generators have been designed 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 important to identify the approximate size and elevation of steamthe examine generator tube leaks and ,to metallographically can be used tosubheader analyze any tube material because this information trend or generic cause of tube leaks.

Conclusive identification of the cause of a steam generator tube leak may enable modifications and/or changes in operation to increase the reliability and life of the steam generators and to prevent a quantity of tube failures in l excess of those analyzed in the FSAR. '

Because of the subheader designs leading to the steam generator tube evaluation of bundles, internal or external inspection and not practical. a tube leak to establish a conclusive cause is Metallographic examination of the accessible i connecting subheader tube will show the condition of the ,

internal subheader wall, giving an indication of the  ;

conditions of '

demonstrating the the leaking tube effectiveness internal wall, thereby of water chemistry controls.  !

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

The surveillance plan outlined above is considered adequate to evaluate steam generator tube integrity and ensure that consequences analyzed in the of postulated tube leaks remain within the lim: tsphe FSAR.

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. o Amendment Na.

Page 3/4 6-QQ .

PCRV AND CONFINEMENT SYSTEMS i 9, 3/4.6.2 REACTOR PLANT COOLING WATER /PCRV LINER COOLING SYSTEM -

OPERATING LIMITING CON 0! TION FOR OPERATION 3.6.2.1 The Reactor Plant Cooling Water (RPCW)/PCRV Liner Cooling System (LCS) shall be OPERABLE with:

4. 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 g ,-

cavity liner bottom head shall be OPERATING; 1

c. At least five (5) out of any six (6) adjacent tubes on *

- the PCRV cavity liner top head and core support floor

- . top casing shall be OPERATING, and

• ~

d. Tubts adjacent to a non-operating tube shall be OPERATING APPLICABILITY: POWER, LOW POWER STARTUP" and SF.700WN' ACTICN

a. With only one (1) RPCW/PCRV LCS looo OPERATING, ensure both heat exchangers are OPERATING in the OPERATING loop, restore the second loop to OPERATING within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in SHUT 00WN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and suspend all operations involving oositive reactivity changes. Without both heat exchangers in the OPERATING loop OPERATING or without any liner cooling system loop flow be in SHUTDOWN within 15 minutes and suspend all operations involving ---^- ' ;f ;. r - --- d 7 '- t positive reactivity changes.

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

. .. - .- =

1 .

Amendment No. '

Page_3/4 6-i DRbr IEB 2 81986

! b. With less thah the above required number of PCRV Liner .

Cooling System tubes OPERATING, other than as in ACTION

a. above, restore the required tubes to OPERATING status l

within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in SHUT 00WN within the following l

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and suspend all operations involving positive reactivity changes.

SURVE!LLANCE REQUIREMENTS 4.6.2.1 The RPCW/PCRV Liner Csoling System shall be demonstrated OPERABLE

.

4. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, by verifying that each PCRV Liner Cooling System loop is circulating cooling water at a flow rate greater than 1100 gpm

b. At least once per 31 days h ifying that liner ooling tube ousies temperature readings and their respective inlet header temperatures (for an operating loop) are within one of the following Ilmits:

4.3) 30 degrees F temperature rise for tubes cooling top head penetrations;

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

/. Exceptions

• I

/) Core Outlet Thermometer Penetrations Tube Delta T 7593 23 degrees F t

/) Core Barrel Seal / Core Support Floor Area Tube Delta T F12T46 47 degrees F F7T43 39 degrees F >

F6T44 43 degrees F i F11T45 38 degrees F F5T47 46 degrees F l

Amendment No.

Page 3/4 6-

• 1 DRA..c .-

/)PeripheralSeal FE8 2 81935 .

Tube Delta T 359 23 degrees F 4S188 23 degrees F 4S10 23 degrees F 35187 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 an instrument failure (4 tubes total) are within their respective temperature Ifmits as specified

• *** ~

3 6 6 af 3 93

c. At least once per "CF"C'.:"^ CTCj oy:

f' Performing a LCS redistribute mode functional test to verify the cooling water to capability of rerouting most of the head. the upper side walls and the top t

/ Performing a functional test to verify the capability to increase the PCRV surge tank pressure to 30 psig by adding helium to the tank.

2.

system flow scanner, associated alarm subheader flow meters.

2, F on e t i.n sil y I e 5T ;*3 T. h < PcRV c_oog g -f 5 T , m fl,w ,eswner 112"'"5 n3

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. 1 Amendment No.

Page 3/4 6-4 DRan,-

PCRV AND CONFINEMENT SYSTEMS FEB 2 e "*:cca 3/4.6.2 REACTOR PLANT COOLING WATER SHUTOOWN /PCRV LINER COOLING SYSTEM _

LIMITING CONDITIONS FOR OPERATIONS-3.6.2.2 The Reactor Plant System

' OPERATING (LCS) with shall be OPERABLE with one loop RPCW/PCRV L each loop OPERATING.at least one heat exchanger and one pump in APPLICABILITY: STARTUP SHUTDOWN [*and REFUELING #

ACTION: With no RPCW/PCRV LCS loop OPERATING, restore at least one ioop CORE to OPERATING status prior to reaching a CALCULATED BULK involving CORE ALTERATIONS-"or ~ -*-^1TEMPEPATURE of 76 in positive reactivity changes. --" resul ting SURVEILLANCE REQUIREMENTS Qg 3n y e veloTi N

4.6.2.2 identified No additional surveillance requirements other than those per Specification 4.6.2.1.

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

(

?OR  :

INFO ONLY

C- =

Aundment No.

Page 3/4 6-D&,dfv BASIS FOR SPECIFICATION LCO 3.6.2 / SR 4.6.2 b 003 Ouring operation at ,

are required to maintain PCRV liner coolingPOWER, and system temperatures two PCRV lin Thermal Barrier andstresses within the FSAR design limits (FSARTSection W.S qpg, is 5.9.2 Evaluation). Liner Cooling Ssystem Design and Design Analytical calculations in support of the PCRV Liner Cooling System design (FSAR Section 5.9.2.4) demonstrata k.T that operation satisfies the criterion whichat full power with one cooling loop forr38 hourD g gy p .r'tt/

increase of 20 degrees F specifies a maximum temperature f concrete. in the bulk temperature of the PCRV f F.'$ A S circulation Operation on one. loop during a loss of forced '

b ' 7 i t t . n .g accident using a PCRV liner cooldown with an increased may resultliner cooling water system cover pressure of 30 of F). 240 degrees F (outlet temperature of approximately 3 analyzedThese conditions condition result in acceptable liner cooling for this and PCRV structural integrity is preserved (FSAR Section 0.1.2.1.5).

The liner local F.

concrete temperatures adjacent to thedegrees However, liner to 150 coolin and their limits follow. potential

• failures of cooling tubes were analyzed PCRV liner or blocking, do not affect the integrity of thelong PCRV as ascooling t such a failure 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 single tube in any adjacent set of six tubes on the PCRV cavity liner top head and core support floor top casing A failed tube which 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),

(FSAR an allowable and acceptable concrete temperature 5.9.2.3.).

Operation hot disclosed of the PCRV spots linerliner.

on the cooling system during startup testing identified These locations were and analyzed engineering evaluation in the above FSAR Sections.

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

Amendment Na.

Page 3/4 6-

.OjY The ACTION times rtB g g ;gy comes from analyses described in FSAR f.e., Section S.9.2.4,sp 48

' hours operation concrete would rise on one loop 20 degrees F. before temperature of the bulk tubes With the number of cooling identify and restore the tube to operating (if possible) statusless than or SHUT 00WN to make permanent repairs.

The to verify operability of the linersurveillance(s) cooling system.

and their respective inte and features Components not safety related do not affect Lcsof the reactor plant cooling water system program a= r=M i n_v. The fet/ P that separate safetyat Fort St. Vrain6erifies OPERABILITYlf thosec.12dh barriers system.

A 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and non safety related portions of the y additional verification of flow as processsurveillance on system flow T. V. rates s pr continuously alarms monitor flow failures would be in each liner cooling loop. Individual tube SURVEILLANCE INTERVAL expected to occur slowly, thus a 31 day d , w e. ,I corrective action. will detect tube failures in time to take With CALCULATED degrees without F, single one operating liner cooling system acceptable loop isBULK C reactivity condition of the reactor and the cooling requirements.,

limited corefailure J ,,IeM When the pCRV pressure INLET TEMPERATURE is less than 200thedegrees core support F,is less than 150 ps i 8" floor zones of the liner concrete temperatures will be cooling system may be valved out as

! Pd V ' g limitation. less than the 250 degree FSAR repairs will not contribute to potentialThus, leaking liner cooling tube the primary system. moisture ingress into l In are determined by thermocouple readings. Surveillance Re Instrument failure (i.e., In the event of an J a failed), the tube with the failed thermocouple may be thermocouple is thought to be considered OPERABLE if thermocouple readings for two adjacent tubes on i

either limits. side of that tube are within their respective temperature thena the Thus, temperature of adjacent tubesrise. would be expec I failure. Power operation may continue untilfailed thermocouple c 1

thermocouple such time as the four adjacent tubes (two on either sidecan be repaired or replaced as long failed of the tube with ture the limits. instrument) are within their respective tempera

Amendment No.

Page 3/4 6-dEy' The use of FFQ n c-division between the APPLICABILITY of 3.6.2.1 Specification 760 deg 3.6.2.2 is explained as follows: and In the FSV HTGR, inlet temperature greater than 760the limiting parameter of interest is a core degrees F. The CALCULATED i

BULK CORE maximum potential TEMPERATURE temperature in the core is a and conservative calculation o components. surrounding The conservatisms BULK CORE TEMPERATURE is limited to 760are degrees F, thesuch that if the CALCUL design inlet temperature of 760 degrees F is not exceeded. Systems used for accident prevention and mitigation are the required to satisfy single failure criterion whenever CALCULATED BULK CORE TEMPERATURE is greate r than 760 degrees F. However, when CALCULATED BULK CORE TEMPERATURE degrees F, it is acceptable to require only oneis OPERABLE equal to orsystem less than 760 required for accident prevention and mitigation as acceptable without single core cooling failure consideration, on the basis of the Ifmited requirements.

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

if forced circulation is temporarily restored, a recalculation can be performed as required based on present plant -

conditions, to establish a new time period CORE TEMPERATURE to reach 760 degrees F. for CALCULATED BULK Redundant systems may also testingbe takenthat provided out of service forced circulationfor maintenance or surveillance is maintained. The time may be recalculated as often as required.to reachF CALCULATED 9

Am;ndment N3.

Page 3/4 6-PCRV AND CONFINEMENT SYSTEMS OSAFT 3/4.6.3 REACTOR PLANT COOLING WATER /PCRV LINER

< TEMPERATURES LIMITING CONDITIONS FOR OPERATION 3.6.3 The RPCW/PCRV Liner Cooling System (LCS) temperatures shall be maintained within the following ifmits:

a.

The maximum average temperature difference between the common PCRV cooling water discharge temperature and the PCRV external concrete surface temperature shall not

• i exceed 50 degrees F.

b.

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

c.

s .

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

d.

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

The than or minimum equal to average LCS 100 degrees F. water temperature shall be greater APPLICABILITY: At all times

! ACTION:

I With any of the above limits not satisfied, restore the limit (s) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or be in SHUTOOWN or REFUELING within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and suspend all operations l

y in positive reactivity changes, or movement of IRR FUEL.

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An.endment No.

Page 3/4 6-DRhy.- '

SURVEllLANCE REQUIREMENTS 4.6.3 The RPCW/PCRV Liner Cooling System temperatures shall be demonstrated to be within their respective limits! least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by: J f./. Verifying that the maximum temperature difference averaged over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period between the PCRV external concrete surface temperature and the common PCRV cooling water discharge degrees F.

temperature in each loop does not exceed 50 2,M.

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

3 /. 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 N 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.

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

y /. 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.

b. AT lent once- Per ~5 ) .J .2 7, L y -fow d i. w a l l y Ter't1 D *- g s s . c . a T e ol P. P e w T e % peraTor e 3

sc ow n t e J l a r m s, c' At least once per 366 days by perfoming a CHANNEL CALIBRATION 4

of the PCRV cooling system temperature scanner, assocfated alams.

seven (97) liner cooling tube outlet thermocouples, an surface temperature indicators.

4 y g A [1 cA *( *Amsndment"* .T 21No. ;*a>

1,1, . t

, . .t->t-es 3 '-

( ;3 o m < " P 2 y* [" au .r,,a<J.' DRAFT FEB 2 e seg BASIS FOR SPECIFICATION LCO 3.6.3/ SR 4.6.3 The temperature limits associated with the Liner Coa'i23'5 System are not soecifically discussed in the FS Various FSARsectionsincluding5.7,5.9,5.Iz,anui., JAR 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.

• Sinca 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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.

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4 Am2ndment No.

Page 3/4.7-  ;

ORA,ey l

PLANT AND SAFE SHUT 00WN COOLING SUPPORT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES - OPERATING 7,- LIMITING CONDITION FOR OPE'.ATION

-- r 3

3.7.1.$ a. The steam valves (V-2214, V-2215, Y-2216, V-2245, generator V-2246, V-2247,superheate . V-2225 and V-2262) shall be OPERABLE with set points in accordance with Table 4.7.1-1, and H

s

b. The. provisions of Specification 3.0./arenotapplicable until 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after reaching 25% RATED THERMAL POWER, to allow testing of the steam generator superheater and reheater safety valves required following. maintenance Surveiflance Requirements or per 4.7.1.5. identified in Specification APPLICABILITY: POWER, LOW POWER, and STARTUP-i ACTION: '

., s With one ' of the required safety valves inoperable, restore the required valve 'to OPERABLE status within

, plant operation"as follows: 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or restrict s

1. W

. ith an EES safety valve inoperable, reduce THERMAL POWER to

, less' ths 50% of RATED THERMAL POWER.

x x - 2. With ~an EES safety valve inoperable while in STARTUP, 3

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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

I Fog INp s .

ON;ov .

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e- -

Amendm:nt No.

Page 3/4.7-DRAFT

_ SURVEILLANCE REQUIREMENTS FEB 2 9 m 4.7.1.5 The sa fety 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 .. nd by verifying the lift settings as specified in Table 4.7.1-1.

9 4

0 4

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,. . ' . , , , , --,.-n- _. -

..- - - - . - , , _ . - - - , , - , - . . . , , , .L.-_ .--. , .-- . , , - . , , -

Am:ndment No.

Page 3/4.7-DRAFT FEB 2 819es TABLE 4.7.I'-1 STEAM GENERATOR SAFETY VALVES

. VALVE NUMBER LIFT SETTINGS .

LOOP I V-2214 V-2215 Less than or equal to 2917 psig V-2216 Less than or equal to 2846 psig V-2225 Less thar. or equal to 2774 psig Less than or equal to 1133 psig LOOP II

• V-2245 V-2246 Less than or equal to 2917 psig V-2247 Less than or equal to 2846 psig V-2262 Less than or equal to 2774 psig Less than or equal to 1133 psig 1

1

. - . .- r.-.-- -,- , - -

y _ ._,-,. _, , -,-,.,__7,--,.____-,y y- ,.__. ---_7-g w w , e

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s. * '

Amsndment No.

Page 3/4.7-DRAFT PLANT AND SAFE SHUTOOWN COOLING SUPPORT fE3 2 8 G33 SYSTEM.S 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 OPERATING section(s) of the steam generator shall be OPERABLE with setpoints in accordance with Table 4.7.1-1.

APPLICABILITY: SHUTDOWN and REFUELING 3

ACTION:~

With less than the above required safety valve (s) OPERA 8LE, restore the required safety valve (s) to OPERABLE status prior F or suspend all operationsto reaching a. CALCULATED BU 2 ::t - ;I involving CORE ALTERATIONS or

<" e . d -^" ::As resulting in positive reactivity

. , . . . , m .g SURVEILLANCE REQUIREMENTS 4.7.1.6 No additional surveillances per Specification 4.7.1.5. required beyond those identified Cf3vi1 0 sI ,?..: ! C r d

3 (a :>*>!! t. \/

Amerdm nt No.

Page 3/4.7-DRAFT BASIS FOR FS 2 81986 SPECIFICATIONS LCO 3.7.1.5/SR 3.7.1.6/SR 4.7.1.6 4. 7.1. 5 AND LCO The economizer evaporator superheater (EES) section of each ste each with one-third nominal relieving capacity The o

reheater section of each steam generator loop is overpressure protected from transients by a single safety valve. These steam generator 10.2.5.3. safety valves are described in the FSAR, Section The above valves are required to be tested in accordance with (ASME Section XI, IGV maintenance. requirements) every 5 years or after tested with steam. To satisfy the testing criteria, the valves must be in Since these valves are permanently installed steam piping, the appropriate means for testing require plant power to be in excess of 22% RATED THERMAL POWER.

' must be conducted during LOW POWER. Thus the test as to minimize operation at power until theConditions are spe,cified s 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 operation is restricted to a condition forone EES safety valve inoperable which the remaining safety valves have sufficient relieving capability to prevent l ove'rpressurization of any steam generator section (i.e.,

boiler feed pump per operating loop). one reheater safety valve inoperable, plant operation Conversely, with any to a more restrictive Mode.

• is restricted A 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> safety in valves ensures that these valves are returnedaction a relatively to service short period of time, during which an overpressure transient is unlikely. Operation at pcwer fcr 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> does not for any extended period. result in a significant loss of safety function The setpoints are those valves identified in the FSAR withfor the safety valve tolerances applied such that the Technical Specifications incorporate an upper bound setpoint. This is consistent with not incorporating normal operating limits in these Specifications.

l

t Ercieru ra 3 :

Table (Page 1 of 3)

Categorization of NRC Comments on TSUP Draft Specifications for Systems, Subsystems and Components with Safety-Related Cooling Functions NRC Comment Specification Number Category Rationale (if needed)

DEFINITION 1.34 1 D Imprecise definition F FSAR clarifications are outside TSUP scope but represent potential RAI.a Table 1.1 1 D Imprecise definition 2 D Imprecise definition Section 3/4.2 1 D or Fb Completeness: FSAR limit is not addressed by Tech Specs.

2 D or F Completeness: FSAR limit is not addressed by Tech Specs.

Section 3/4.4 1 D or F Completeness: FSAR assumed conditions are not confirmed by a LCO compliance.

2 D or F Completeness: FSAR assumed conditions are not confirmed by a LCO compliance.

3 D or F Completeness: FSAR assumed conditions are not confirmed by a LCO compliance.

4 D or F Clarity / Completeness:

An apparent FSAR assumed condition is not confirmed by a LCO compliance.

5 F or D FSAR clarifications are outside TSUP scope unless used specifically to limit TSUP coverage. Issue is still a possible RAI.

6 D Clarity Section 3/4.5 1 D Clarity LCO 3.5.1.1 1 D 2 D Clarity 3 D 4 D a

RAI: Request for Additional Information Pursuant to 10 CFR 50.54(f).

b Within the TSUP scope, the issue of completeness must be discussed furth r with regard to possibly requiring new or additional Tech Specs as opposed to merely ensuring the adequacy of existing Tech Specs.

~

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e

• Table (Page 2 of 3)

NRC Comment Specification Number Category Rationale (if needed)

LCO 3.5.1.1 5 D (continued) 6 D 7 D 8 D 9 D 10 D 11 D

! 12 D 13 D .

14 D

. 15 D 16 D 17 D Clarity / Completeness 18 D 19 D Clarity / Completeness 20 D 21 D 22 D .

23 D Basis needs to be revised by deletion or clarification.

, F FSAR clarifications are outside TSUP scope but represent-potential RAI.

24 .D, Basis to be revised i

for clarity.

F FSAR clarifications are outside TSUP scope but represent potential RAI.

LCO 3.5.1.2 1 D 2 D 3 D Clarity 4 D*

5 D*

LCO 3. 6.2.1 1 D FSAR does not clearly support LCO time allowance.

3 D 4 D 5 D LCO 3.6.2.2 1 D 2 D*

3 D I

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Table (Page 3 of 3)

NRC Comment Specification Number Category Rationale (if needed) 4

LCO 3.6.3 1 D*

2 D 3 D FSAR does not clearly support Tech Spec limit.

3 Section 3/4.7 1 D Clarity j 2 D or F Completeness: FSAR assumed conditions are not confirmed by a LCO compliance.

LCO 3.7.1.1 1 D See comment 24, LCO 3.5.1.1

F See comment 24, LCO 3.5.1.1 2 D Clarity

3 D Clarity .

4 F FSAR clarifications are.

)j outside TSUP scope but represent potential RAI.

5 B .

1 LCO 3.7.1.2 1 D Clarity / Completeness

! LCO 3.7.1.5 1 D 1- LCO 3.7.1.6 1 D*

J

, LCO 3.7.3 1 D Clarity i

LCO 3.7.5 1 D Clarity j' LCO 3.7.8 1 D Clarity 4 D Clarity / Completeness 3 F FSAR clarifications are

outside TSUP scope but represent potential RAI.  ;

1 4 B '

a LCO 3.9.1 1 D or F Completeness / Clarity; -

Apparent FSAR assumed f -

conditions are not confirmed by either definition of operating

, mode or a LCO compliance.

F FSAR clarifications are outside TSUP scope but l represent ' potential RAI.  !

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