Text

Forwards Request for Addl Info Re Tech Spec Upgrade Program, Including Addition of Calibr Requirement for Seismic Instruments Determined to Be Out of Calibr Following Seismic Event
	ML20235F336
Person / Time
Site:	Fort Saint Vrain
Issue date:	07/02/1987
From:	Heitner K; Office of Nuclear Reactor Regulation
To:	Robert Williams; PUBLIC SERVICE CO. OF COLORADO
References
	TAC-56565, NUDOCS 8707130334
	Download: ML20235F336 (43)
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UNITED STATES i

NUCLEAR REGULATORY COMMISSION 4

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,E July 2,1987 j# J I

Docket No. 50-267 Mr. R. O. Williams, Jr.

Vice President, Nuclear Operations Public Service Company of Colorado Post Office Box 840 i

Denver, Colorado 80201-0840 j

Dear Mr. Williams:

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SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR TECHNICAL SPECIFICATION UPGRADE PROGRAM FOR FORT ST. VRAIN NUCLEAR GENERATING STATION (TAC NO. 56565) <

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References:

See Enclosure 1 We have reviewed the information you submitted with your letter, Reference 1, on the Technical Specification Upgrade Program for the Fort St. Vrain Nuclear Generating Station (FSV). Your submittal, provided documentation of comment resolutions that were reached between Public Service Company of Colorado (PSC) and the Nuclear Regulatory Commission (NRC) during a meeting on October 27-30, j 1986, regarding the FSV Technical Specification Upgrade Program (TSUP). This l meeting was held to resolve comments provided by the NRC in Reference 2, on the November 30, 1985 FSV TSUP draft submittal. The NRC provided a meeting summary in Reference 3, which tabulates the comments discussed and the nature of their resolution.

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This letter provides a request for additional information (See Enclosure 2) as a result of our review of your current submittal, against our original comments (Reference 2). The comments are identified in the same manner as used in our meeting summary (Reference 3). Also, in Enclosure 7, we have provided requested actions for PSC resulting from our review and resolution of the eleven NRC action items from the meeting of October 27-30, 1986, and for the action items summarized in Enclosure 1 to Reference 2 that still require further PSC action.

Further, Enclosures 3 and 4 contain supplements to our comments transmitted in Reference 4 on the safety-related cooling functions. These supplemental comments address our view of how your proposed revision to Technical Specifica-tion LCOs 4.3.1-(Reference 6) and 4.1.9 (Reference 7) impacts the review of the final draft of the FSV TSUP. This is in response to your desire to have the reviewers of the TSUP final draft satisfied that your proposed LC0 4.1.9 resolves any previous comments or concerns raised during the TSUP final draft review. Enclosure 3 provides the supplemental concerns and their proposed resolutions. Enclosure 4 is our markup of the November 30, 1985 draft Technical Specifications that incorporate our proposed resolutions.

8707130334 870702 l PDR ADOCK 05000267 p PDR

b e Enclosure 5 transmits our proposed categorization of our comments on the Ausniary Electrical Power Systems, Section 3/4.8, as transmitted to you in i Reference 5. This proposed categorization uses the same designations as used !

in previous correspondence, for example, Reference 3. This proposed categoriza-tion is for your information and use in preparing your response to these i comments.

Also, PSC should consider transmitting in TSUP format all Fort St. Vrain 4 Technical Specification Amendments that have been issued in parallel with, but not included in the TSUP final draft of November 30, 1985, and any that may be issued in the future, but prior to the TSUP amendment application. As a f l minimum this would include Amendment Nos. 47 through 55. Receipt of the j material in the TSUP format will expedite our review and transmittal of any i comments prior to this material being submitted in the TSUP amendment application. j Finally, our review of your TSUP amendment application will be expedited further if you would transmit your proposed resolutions to the forty-two "A#"

(PSC action items) and twenty-four "C" (PSC to explain in proposed safety I evaluation) Category items (See the list of these items in the October 27-30, !

1986 meeting summary, Reference 4). Again, this will allow us the opportunity j to review and comment on these resolutions prior to their being submitted in '

the TSUP amendment application.

Please provide the required information within 60 days of receipt of this letter. If you feel that further discussion would be helpful in resolving these open issues, please call me at (301) 492-7592.

The information requested in this letter affects fewer than 10 respondents; therefore, OMB clearance is not required under P.L.96-511.

Sincerely, '

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Kenneth L. Heitner, Project Manager Project Directorate - IV Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation

Enclosures:

As stated cc w/ enclosures:

See next page

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DISTRIBUTION 1)ocket File NRC PDR Local PDR E. Luce PD4 Reading D. Crutchfield F. Schroeder J. CalvoP. Noonan K. Heitner OGC-Bethesda E. Jordan J. Partlow ACRS (10)

J. Miller , R. Emch y PD4 ant File PD4/LA PNoonan hN PD4/PM KHeitner:sr h PD /D JCalvo ( er TSB REmc g

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. p s- Mr. R. O. Williams Public Service Company of Colorado Fort St. Vrain cc:

Mr. D. W. Warembourg, Manager Albert J. Hazle, Director Nuclear Engineering Division Radiation Control Division Public Service Company Department of Health of Colorado 4210 East lith Avenue P. O. Box 840 Denver, Colorado 80220 Denver, Colorado 80201 Mr. David Alberstein, 14/159A Mr. R. O. Williams, Acting Manager GA Technologies, Inc. Nuclear Production Division j Post Office Box 85608 Public Service Company of Colorado i San Diego, California 92138 16805 Weld County Road 19-1/2 ls Platteville, Colorado 80651 Mr. H. L. Brey, Manager l Nuclear Licensing and Fuel Division Mr. P. F. Tomlinson, Manager Public Service Company of Colorado Quality Assurance Division P. O. Box 840 Public Service Company of Colorado Denver, Colorado 80201 16805 Weld County Road 19-1/2 j Platteville, Colorado 80651 -

Senior Resident Inspector U.S. Nuclear Regulatory Commission Mr. R. F. Walker !

P. 0. Box 840 Public Service Company of Colorado j Platteville, Colorado 80651 Post Office Box 840 j Denver, Colorado 80201-0840 ;

Kelley, Stansfield & 0'Donnell j Public Service Company Building Commitment Control Program Room 900 Coordinator 550 15th Street Public Service Company of Colorado i Denver, Colorado 80202 2420 W. 26th Ave. Suite 100-D 6 Denver, Colorado 80211 -

Regional Administrator, Region IV ] '

U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Chairman, Board of County Commissioners of Weld County, Colorado I Greeley, Colorado 80631 Regional Representative Radiation Programs Environmental Protection Agency 1 Denver Place 999 18th Street, Suite 1300 Denver, Colorado 80202-2413 1

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, ENCLOSURE 1 REFERENCES

1. H. L. Brey letter to H. N. Berkow, Technical Specification Upgrade Program, February 20,.1987, Public Service Company of Colorado, (P-87063).

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

3. Memorandum for H. N. Berkow from C. S. Hinson, Summary of October 27-30, 1986 Meeting at Fort St. Vrain (FSV) to Discuss Staff Comments on the FSV Technical Specification Upgrade Program (TSUP), December 15, 1986, U.S.

Nuclear Regulatory Commission.

.4. K. L. Heitner letter to R. O. Williams, Jr. , NRC Comments on the Technical Specification Upgrade Program (TSUP), LCOs for Safety-Related Cooling

!- Function, April 17, 1987, U.S. Nuclear Regulatory Commission.

5. K. L. Heitner letter to R. O. Williams, Jr. , Draft Updated Technical l Specification 3/4.8, Request for Additional Information (TAC No. 56565),

May E,1987, U.S. Nuclear Regulatory Commission. ;

6. R. O. Williams letter to H. N. Berkow, Proposed Technical Specification Change Eliminating Reliance on Reheater Section of Steam Generator for Safe Shutdown Cooling, January 15, 1987, Public Service Company of Colorado,(P-87002).

7. Draft R. O. Williams letter to H. N. Berkow, Proposed Technical Specifi-cation Change, LCO 4.1.9, Core Inlet Orifice Valves / Minimum Helium Flow and Maximum Core Region Temperature Rise, April 17, 1987, Public Service Company of Colorado, (P-87124).

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' ENCLOSURE 2 ADDITIONAL INFORMATION NEEDED TO COMPLETE REVIEW 0F TECHNICAL SPECIFICATION UPGRADE PROGRAM FOR FORT ST. VRAIN NUCLEAR GENERATING STATION The following is a list of requests for additional information needed to complete this review. The identification of each item is the same as in the NRC notes of the October 27-30, 1986 meeting (G. L. Plumlee Memorandum Dated, December 1,1986) unless noted otherwise.

I TABLE 1.0-1 The Licensee should provide additional justification for their proposed mark-up of Table 1.1 (P. 1-9, Attachment 1 of PSC letter of February 20, 1987) or should make the "0" and *#" footnotes specific to the precise evolutions (tests) or surveillance involved. PSC references such provisions in the GE-BWR STS. However, neither the GE-BWR STS, NUREG-0123, Rev. 3, nor the Perry TS (example given out in October 27-30, 1986 nmeting by PSC) allow indiscriminate switching as would be allowed with the PSC proposed words of "...for the purpose of performing surveillance or other tests...". Both the GE-BWR STS and the Perry TS allow switching of the Mode Switch Position only under rather precisely defined conditions, such as "...to test the switch interlock functions.. ." or ".. .while a single control rod drive is being removed from the reactor pressure vessel per Specification 3.9.10.1. .." or ". . .while a single control rod is being recoupled or withdrawn provided that the one-rod-out interlock is OPERABLE".

Further, please revise the pound sign (#) footnote in Table 1.1 by deleting the words "provided that k,77 is verified less than 0.99" and replacing the deleted words with "provided that the control rods are verified to remain fully inserted." As acknowledged in the PSC response to NRC Comment No. 4 on TSUP draf t LCO 3.1.3, the use of the word " verified" is incorrect in establishing a quantitative estimate of SHUTOOWN MARGIN (that is, keff) because the reactor operators cannot actually verify the accuracy of the calculated assessments provided to them. However, 1

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g s consistent with the equivalent footnote provided in Table 1.2 of the BedR-STS (NUREG-0123, Rev. 3), the operators can verify that control rods !

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SL 2.1.1-5 The Licensee should provide additional information and a safety evaluation to support splitting the existing fSV Safety Limit 3.1, part into the proposed Safety Limit Section 2.1.1 (November 1985 Draf t) and part f

into a Limiting Condition for Operation LCO 3.2.6 (November 1985 Draf t).

The 24 hr action time has been adequately addressed in PSC's letter of February 20, 1987. However, the reference to FSAR Revision 4 Section 3.6.8 does not provide specific justification for downgrading part of the existing Safety Limit to a Limiting Condition for Operation. In fact, fSAR Revision 4 Section 3.6.8 titled " Core Safety Limit" discusses all of the limits in the proposed SL 2.1.1 and LCO 3.2.6 as if they were safety limits as does the existing FSV SL 3.1.

LC0 3.0 - New Item Please provide an equivalent LC0 to the recently proposed LCD 4.0.4 in l the existing Technical Specifications.

l LC0 3.1.1-1 Please revise the wording of TSUP draft LCO 3.1.1.a to add the words "from the fully withdrawn position" immediately after the words "152 seconds." PSC was to propose an alternative to the May 30, 1986, NRC markup of the TSUP draft. The proposed revisions to the basis does not adequately clarify the limiting condition for operation. In addition, the wording of the proposed insert at the bottom of page 3/4 1-5 in the PSC markup (Attachment 1 to P-87063) needs to be revised to read as follows:

The full insertion scram time can be determined either directly from a full insertion scram time or indirectly from a partial scram time of 10 inches or more. For the partial scram time, the estimate of an 1

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' extrapolated scram time is always based on assuming scram from the fully withdrawn position and not from the~ actual rod position.

LCO 3.1.3-4,-8 i

Please provide additional information with regard to the PSC position on the substantiation and verification of nuclear methods as expressed in the PSC response to NRC Consnent Nos. 4 and 8 that are given in Attachment 2' to P-87063. Is it PSC's position that the verification and documentation-of the assessment methodology used in LCO 3.1.3 meet the intent, if not all the specific requirements or guidelines, of industry standards per ANSI /ASME N45.2.11, ANSI /ANS-19.3 (Section 6), ANSI /ANS-19.4, ANSI /ANS-19.5, and ANSI /ANS-19.6.17 Are detailed records or other supporting documentation for the PSC position maintained from the reviews j of the base reactivity curve 'as required to be performed by the Nuclear I facility Safety Committee (NFSC) per the existing Technical Specifications LCO 4.1.8 and SR 5.1.47 Can it be demonstrated from the NFSC records that the type of analytical-to-experimental comparisons cited in the PSC position have been and are being factored into the NfSC review? If not, explain how the NFSC has verified the continued use of the base. reactivity curve to assure an adequate SHUTDOWN MARGIN per the basis of existing LCO 4.1.8 without documenting and updating such comparisons as those cited in the PSC position? If such documentation is not being maintained, proposed TSUP draft Specification AC 6.5.2.7.1 needs to be changed such l that the NFSC 1,s specifically charged with reviewing and documenting the results of the performance of SR 4.1.3, SR 4.1.4.1.2, SR 4.1.5, and j SR 4.1.7 and the results of comparing the reload design against the requirements specified in DESIGN FEATURE 5.3.4. The reviews and documentation are both necessary and appropriate so that there is a record i of the verification process for assuring the adequacy of the SHUTOOWN MARGIN assessment methodology employed in LCO 3.1.3 and SR 4.1.3. This approach is consistent with the intended role of the NFSC as agreed upon with NRC and acknowledged by PSC in the Attachment to PSC letter, C. K. Millen to R. S. Boyd, September 11, 1975.

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LCO 3.1.6 New Item (Old LCO 3.9.1-8) 1 The Licensee should add SHUTDOWN and REFUELING to the Applicability j statement of LC03.1.6 on the Reserve Shutdown System. Such added l appitcability should account for the exception required for any two control rod pairs which may be removed f rom the PCRV (LC0 3.1.4.2.a.1 of l November 30, 1985 Draft). Although the May 30, 1986 NRC connent l

LC0 3.9.1-8 on inadequate SHUTDOWN MARGIN during REFUELING was categorized as an "F" (futher discussion possible) in the October 1-2, 1966 meeting, further comparison to the GE-BWR STS Rev. 3, NUREG-0123, indicate that acceptance of small SHUTDOWN MARGINS (LCO 3.1.1 P 3/41-1) during REFUELING with any control rod withdrawn is with the proviso that the Standby Liquid j Control System be OPERABLE (LCO 3.1.5, P 3/41-19). The GE-BWR STS is used !

f or comparison rather than the W-STS as the f t. St. Vrain reactivity control system of control rods and reserve shutdown material is similar to l

the GE reactivity control system of control rods and Standby Liquid Control ,

i System, with neither having routine boration reactivity control as in the !

Westinghouse system. Also, for f t. St. Vrain, the Actions b.2 for SHUTDOWN and C.2.b f or REFUELING with inadequate SHUTDOWN MARGIN (LCO 3.1.3 of November 30, 1985 Draf t) require actuation of sufficient reserve shutdown material to achieve the required SHUTDOWN MARGIN. Yet in the Reserve l Shutdown System Specification (LCO 3.1.6 of the November 30, 1985 draft) ,

there is no requirement for OPERABILITY in either the SHUTDOWN or REFUELING ;

condition.

LC0 3.2.1-1 Please expand the Basis to describe briefly what activity is involved in the " determination by evaluation." What reactor observables are evaluated?

LCO 3.3.2.3-3 The Licensee should add a calibration requirement in SR 4.3.2.3.2 for seismic instruments determined to be out of calibration following a seismic event. As indicated in the basis, P. 3/4 3-83 of the November 30, 1985 4

. o Draf t, the Licensee has already consnitted to this (last sentence, fourth paragraph). However, there is presently no requirement for such calibration in the surveillance. Although the STS guidance is to do such calibration within ten days, it is judged that 30 days provides for allowing monitoring of after shocks following the initial event. I Experience with other commercial plants is that the contractor / manufacturer provides on-site calibration capability. PSC should investigate such on-site service capability from their seismic instrument manufacturer so that the seismic instruments remain on-site to be maximum 1y available for monitoring af ter shocks.

LC0 3.3.2.3-4 The Licensee should specify monthly CHANNEL CHECKS for the seismic instruments in TS Table 4.3.2-2 (November 1985 Draf t, P. 3/4 3-82) rather than the proposed quarterly checks. PSC had proposed quarterly CHANNEL CHECKS as consistent with Turkey Point seismic instrument surveillance.

However, as Turkey Point specifications are the exception to the rule and as Turkey Point is situated in Seismic Zone 0 and FSV is situated in Seismic Zone 1, the Turkey Point specifications are not a reasonable comparable to use. Additionally, it is in the Licensee's interest to have the seismic instruments operable to facilitate restart if shutdown occurred. The relatively uncomplicated CHANNEL CHECKS would enhance )

instrument operability and thus the assessment of a seismic disturbance and the implications to plant restart capability.

t LC0 3.5.4 - New Item Please revise LCO 3.5.4 ACTION b (second part on page 3/4.5-30 of the TSUP draft). The words, " establish a backup system for fire suppression purposes within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ," need to be changed to the wording, " establish a backup system for fire suppression purposes prior to reaching a CALCULATED l

BULK CORE TEMPERATURE of 760*f but within a period of time not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ." The revision is necessary to assure an operable flow path to the liner cooling system (LCS) via the firewater system when the LCS is the l only decay heat removal path during an interruption of forced cooling for i

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purposes of maintenance and inspection. This change is consistent with the NRC guidelines provided in the NRC letter of December 5, 1986, on PSC commitments required to approve the proposed version to LC0 4.1.9 in the existing Technical Specifications, q

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LC0 3.6.4-2 l l

Please provide additional information with regard to the source and !

approval of the acceptance criteria for PCRV concrete permeability and PCRV liner thinning as cited in the PSC markup of pages 3/4.6-39 and 3/4.6-40 of j the TSUP draft in Attachment 1 to P-07063. The NRC does not have a copy of the previous ISI procedures alluded to in the citations, and these ,

acceptance criteria are not found in the recent ISIT submittals.

LC0 3.7.2-4 The Licensee should provide additional clarification in the basis on P. 3/4 7-14, third paragraph, second and third sentences, as they read:

1 "A 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ACTION time is provided to isolate the affected loop, in an l effort to regain OPERABILITY of a second hydraulic field pump and/or at j least one accumulator for the affected valve group. If OPERABILITY of a second hydraulic fluid pump and/or at least one accumulator is not restored within i hour, reactor shutdown is required with 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ."

Since OPERABILITY of a second hydraulic fluid pump is not mentioned in ]

the Action Statement, only " supply of at least 2500 psig", the basis should ;

use the terminology of the Action Statement, namely, supply of at least !

2500 psig. Also, the basis has interpreted the Action Statement to allow j one hour to restore the required conditions of at least one accumulator and ;

or at least 2500 psig pressure before reactor shutdown within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The Action Statement a. of P. 3/4 7-12 does not address a restoration time. It requires isolation of the affected loop in one hour and reactor shutdown in the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months without regard to any restoration j time.

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LC0 3.7.6.3-3 l

For the halon system in Building 10, please propose specifications consistent with SR 4.7.6.3 in the WNP No. 2 Technical Specifications, NUREG-1009 (see attached page). At WNP No. 2, tank " quantity" is determined once per six months using the heat tape and gun method approved by the American Nuclear Insurers. However, storage tank weight must be i 1

verified periodically. PSC needs to specify an appropriate surveillance period for verifying storage tank weight and to justify any period exceeding 36 months.

I LCO 3.7.8-4 l Please correct the misspellings in PSC's markup of TSUP draft LC0 3.7.8 ACTION b. The word, " values," is misspelled twice as " valves."

LOC 3.7.10-3 Please revise FSAR Sections 1.4 and B.5.2.7 to be consistent with the Basis definition of " safety-related" as including Class la components. PSC l

expressed the desire to retain the wording in the opening sentence of the Basis as being indicative of their position with regard to the scope of the safety-related snubbers. The Basis and the FSAR reed to be consistent on l l this point.

l LC0 3.9.1-2 l The Licensee should retain the APPLICABILITY as it was in the November 1985 Draft or should provide additional information or changes to split out Specifications 3.9.la and b from Specifications 3.9.lc and d.

l Specifications 3.9.la and b are justly applicable to only "whene'fer both l primary and secondary PCRV closures of any PCRV penetrations are removed" l which is PSC's proposed APPLICABILITY. However, Specifications 3.9.1 c and d on requiring two startup channel neutron flux monitors and l maintaining the SHUTDOWN MARGIN requirements of Specification 3.1.3, are l

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PLANT SYSTDt5 HALON SYSTEMS i

LIMlf!M CONDITION FOR OPERATION 3.7.6.3 The 18 Halon systees In the PGCC units in the control ' room shall be OPERA 8LE with the storage tanks having at least 954 of full charge weight and 901 of full charge pressure.

APPLICA81LITY: Whenever equipment protected by the Halon systems is required to be OPERA 8L(,

ACTION:

a. With one or more of the above required Halen systems inoperable. I within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a continuous fire watch with backte fire suppression equipment for those areas in which redundant systees or

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components could be damaged; for other areas, establish an hourly ffre watch patrol. i I

b. The provisions of Specif fcations 3.0.3 and 3.0.4 are not applicable.

I SURVEILLANCE RE001REMENTS L

4.7.6.3 Each of the above required Halon systems shall be demonstrated OPERA 8LE:

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a. At least once per 31 days by vertfying that each valve (aanual, i power-operated, or automatic) in the flow path fs in its correct position.

b. At least once per 6 months by verifyin0 Halon storage task quantity and pressure,

c. At least once per 18 months by:

1. Verifying the system, including associated ventilation systee i fire dampers and fire door release mechanisms, actuates.

manually and automatically, upon receipt of a simulated 1 actuation sfpsal, and l

2. Performance of a flow test through accessible headers and nozzles to assure no blockage.

d. At least once per 36 months by vertfying Halon storage tank welght.

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4 applicable throughout the REFUELING mode. Therefore, either the original APPLICABILITY in the November 1985 Draft should be retained or 3.9.la and b need separated from 3.9.lc and d with different applicability statements as discussed above. .

LCO 3.9.1-3 The Licensee should revise the wording as suggested for Action b of LC0 3.9.1 (similar wording should also be used for Action C.1 of this LCO) to:

"With one of the above required neutron flux monitors inoperable, or not OPERATING, immediately suspend all operations involving CORE ALTERATIONS, any evolution resulting in positive reactivity changes, or movement of IRRADIATED FUEL.*

The Licensee's position that the proposed words ". . . control rod movements resulting in positive reactivity changes . . .* were intended to be specific versus the NRC recommended wording ". . . any evolutions resulting in positive reactivity changes . . ." has been accepted for the other LCOs involved LC0's 3.6.5.1-2, 3.6.5.2-1, 3.7.1.1-2, and 3.7.9-3 (other NRC actions items from October 27-30, 1986 meeting concerning reactivity changes), for those LC0's, flexibility to change between feedwater and condensate cooldown is accepted because the positive reactivity effect of any involved cooldown has already been accounted for !

in the SHUTDOWN MARGIN requirement of 0.01 delta k in accordance with i LC0 3.1.3. And the reactivity additions are relatively small and added slowly as PSC explained in Attachment 4 to their February 20, 1987 letter. !

for the subject LCO, however, Actions b and and C.) involve loss of one and both startup channel neutron flux monitors, respectively. Allowing an intended positive reactivity change due to cooldown with degraded startup j channel neutron flux monitoring is unacceptable. When both startup {

channels are inoperable, any controllable positive reactivity change should be stopped, as under these conditions, immediate assessment of flux changes is lost. With only one operable startup channel of neutron flux monitoring, it is also unacceptable to intentionally make positive 9

reactivity changes as that one startup channel may be operating erroneously and there is no second operating channel to confirm it's readings. Also, with only one operating channel, a sudden loss of that channel again l results in complete loss of the ability to make immediate assessment, of neutron flux changes. These changes would make the FSV Actions consistent with those of STS Rev 5, P. 3/4 9-2, on neutron flux monitoring capability during ref ueling, and with those of the existing TSV Technical Specifications, LCO 4.7.1. ]

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AC 6.3/6.4-1 The Licensee should include reference to the NRC March 28, 1980 letter in Technical Specification Sections 6.3.2 and 6.4.1 per the subject NRC comment. Contrary to the Licensee's position in the October 27-30, 1986 Meeting that the NRC March 28, 1980 letter is not in their letter log and ,

therefore is doubtful that is was agreed to, the Licensee responded to it in their letter of December 20, 1980 (P-80438). In Attachment 1 of that letter, the Licensee stated that although they had received the NRC l March 28, 1980 letter too late to implement the requirements by August 1,1980 as required by the letter's Enclosure 1, some requirements l would be met by the August 1, 1980 date and their program for compliance l would be submitted by January 15, 1981. As it appears that the Licensee has consnitted to the subject NRC March 28,1980 requirements, stating this in the Technical Specifications should not involve any undue hardships.

AC 6.5.1.6a (AC 6.5-1)

The Licensee should revise AC 6.5.1.6a to require the PORC to review .

any procedures required by AC 6.8.4 as AC .8.4 includes procedures for todine sampling in the reactor building and Post-Accident Sampling, both of which are TMI-2 Action items of NUREG-0737 and Generic Letters 83-36, 37.

Also, the Licensee should revise AC 6.5.2.7 to require the NFSC to review the programs of AC 6.8.4. This is what PSC said PORC and NFSC do now (See PSC letter of February 20, 1987, Attachment 2). However, as written, AC 6.5.1.6a and AC 6.5.2.7 do not have these requirements but should.

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s AC 6.5.1.7b (AC 6.5-1)

The Licensee should provide additional information to explain what

" Procedure Deviation Reports" are, as used in their letter of i druary 20, 1987, Attachment 3, Item AC 6.5# 1-7 and Attachment 1, P. 6-15, marked up item 6.5.1.7.b. In Attachment 3, reference is made to

" Temporary Changes" whereas in Attachment 2, reference is made to-

" Procedure Deviation Reports". It is not clear if these are one and the (

l same thing. If they are the same, PSC should provide additional l

information to justify the proposed exception for " Temporary Changes". ,

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AC 6.9.1.2.a ( AC 6.9-1)

The Licensee should revise their proposed marked up P. 6-27 (PSC lj letter of February 20, 1987, Attachment 1) to either put report submittal times in 6.9.1.2, first paragraph as in the STS Rev. 5, P. 6-16, or place a j report submittal directly in 6.9.1.2.a.2. As proposed, 6.9.1.2.a.2 does not have any report submittal time connected to it. PSC placed their report submittal directly in 6.9.1.2.a.1 rathen than 6.9.1.2. Ther ef or e, when 6.9.1.2.a.2 was added, it was not covered by a report submittal time. I 1

The following RAI is a result of the review of Attachment 1 to P.-85448 NRC Action 3 I

l PSC should provide the Environmental Qualification study for this item.

The following is NRC Action Item 3 and its response from Attachment 1 to PSC letter of November 27,1985 (P.- 85448) .

NRC Action Provide guidance on whether components, which are required to function l to maintain other equipment within an environment for which it is qualified, should also be in the Technical Specifications (for example, main steam isolation valves).

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NRC Response !

Where assumptions for equipment operability related to environmental qualification are based on the successful operation of active components in the event of an accident, the availability and reliability of these l components should be ensured through Technical Specification requirements.

Specific items of concern identified by the staff were the main steam isolation valve (MSIV) operability and closing time requirements as well as the hot reheat (HRH) valve operability requirements. Other EQ operability-l requirements should be identified by the licensee and incorporated as l

Technical Specification requirements to ensure equipment necessary to mitigate accidents function within the assumed environmental conditions.

In addition to the EQ analysis, other analysis which rely on equipment ;

l operability may also result in Technical Specification requirements. For i example, the equipment operability requirements based on previous fire j protection analysis (Appendix R Evaluation: FSV Reports 1 through 4) I should be reflected in the Technical Specifications.

i i

1 12 l

. ENCLOSURE 3 .,

Additional Comments on The Fort St. Vrain Technical Specification Upgrade' Program Final Draft LCOs ;

for Safety Related Cooling Functions NRC COMMENTS-LCO 3.5.1.1 i

1. Based on the PSC letter (P-87002) dated January 15, 1987, the condition statement for LCO 3.5.1.1.a.2 needs to be rewritter, as follows: a Both steam generator sections (both the- economizer-evaporator-superheater (EES) and the reheater) OPERABLE including two OPERABLE flow

-paths.

I

2. Previously, the safe shutdown cooling outlet flow paths were via the i by-pass valves off each loop's .superheater outlet and each loop's hot reheat steam line. The by-pass valves were verified to be OPERABLE as part of the normal operation of the bypass function. The recently installed six inch vent lines described in F-87002 are apparently not to be used on a routine basis. Therefore, SR 4.5.1.1.b needs to be revised by renumbering surveillance b.2 and b.3 to b.3 and b.4, respectively, and adding a 'new surveillance as SR 4.5.1.1.b.2. The new SR 4.5.1.1.b.2 should read as follows:

At least once per 18 months by verifying the OPERABILITY of each superheater outlet flow path by verifying that the valves in the six ,

inch vent lines can be opened and that the vent flow paths are not );

obstructed.

i

3. Subject to NRC final approval of the proposed revisions to the Basis 1 for the existing LCO 4.1.9, the following paragraph needs to be added et the bottora of the fourth page of the Basis for LCO 3.5.1.1 following the second paragraph of the subsection entitled Redundancy Criteria:

Specification 3.0.N provides the methodology and necessary data to determine the appropriate time interval to reach a CALCULATED BULK CORE TEMPERATURE of 760 degrees F. If the active core remains below this temperature, which corresponds to the- design maximum core inlet temperature as indicated above, then the design core inlet temperature cannot be exceeded and there can be no damage to fuel or PCRV internal components regardless of the amount, includin5 total absence, or reversal, of primary coolant helium flow.

PSC needs to provide the appropriate number "N" for the cross referenced specification. The need for Specification 3.0.N has been identified in the NRC Request for Additional Information at Enclosure 2.,

4. The first paragraph of the subsection entitled Steam Generators on the fifth page of the Basis for LCO 3.5.1.1 and LCO 3.5.1.2 needs to be replaced with the following paragraph.

4 ' k 2

Whenever the CALCULATED BULX CORE TEMPERATURE exceeds 760 degrees F, .

both the reheater and EES sections of the steam generator must be OPERABLE. The steam generator reheater or EES sections can receive water from either the emergency condensate header or the emergency feedwater header as required to be OPERABLE per this- Specification and per Specification 3/4.5.3. System flow OPERABILITY is determined by verifying flow from each of the aforementioned emergency headers (see LCO 3.5.3.1) through each section of each steam generator. Whenever the l- CALCULATED BULK CORE TEMPERATURE is less than or equal to 760 degrees ~ F or the plant OPERATIONAL MODE is REFUELING, system flow OPERABILITY is determined by verifying flow from either of the aforementioned emergency headers (see LCO 3.5.3.2) through either section of either steam l generator.

5. An additional paragraph is also required under the subsection entitled Steam Generators on the fifth page of the Basis for LCO 3.5.1.1. This paragraph needs to discuss the appropriate operability requirements for the seismically and environmentally qualified six inch vent lines and to cite the supporting safety analysis requiring the use of these vent lines.

PROPOSED RESOLUTIONS

1. /2 . / 3 . /'4 . /5 . PSC needs to incorporate the required changes.

NRC COMMENTS-LCO 3.5.1.2 1, Contrary to NRC Comment No. 1 on LCO 3.5.1.2 as given in Enclosure 1 to the NRC letter, Heitner to Williams, April 17, 1987, the words to be added after the word "0PERABLE" in the condition statement for LCO 3.5.1.2.a.2 need to be " including one OPERABLE flow path," not two. LCO 3.5.3.2 requires only one flow path, either the emergency condensate header or the emergency feedwater header, to be OPERABLE in STARTUP and 7HUIDOWN whenever the CALCULATED BULK CORE TEMPERATURE is less than or eqe to 760 degrees F or in REFUELING. j

2. See NRC Comment No. 4 on LCO 3.5.1.1.

3. Delete LCO 3.5.1.2.b.1 and renumber b.2 through b.5 as b.1 through b.4, respectively. The deleted LCO is neither needed nor appropriate since the i boiler feed pumps are not required to be OPERABLE under the same APPLICABILITY statement (see LCO 3.7.1.1) nor necessarily the emergency feedwater header (see LCO 3.5.3.2). For decay heat levels below that for shutdown from 35% power, previous analysis has shown that fuel damage will not occur in the depressurized core as long as one train of the PCRV LCS is operating. Necessary modifications to the TSUP draft to incorporate the guidelines from the NRC letter, Heitner to -Williams, dated. December 5, 1986, are indicated elsewhere in these comments, and these modifications are directed to providing assurance that the PCRV LCS provides a backup cooling capability during the conditions of APPLICABILITY. However, NRC l

I reserves the right to modify LCOs 3.5.1.2, 3.5.3.2, and 3.7.1.1 pending completion of the review on the occurrence frequency of rapid depressurization events.

, . 3 l l

l I

In addition, further modifications to Specifications 3/4.5.1, 3/4.5.3, and 3/4.7.1.1 may be required depending upon the final resolution to NRC '

Comment No. 1 on Section 3/4.4 as documented in Enclosure 1 to the NRC letter, Heitner to Williams, April 17, 1987. As noted in the subject comment, the Updated FSAR does not support operation with the reactor )

producing fission heat simultaneously with ' reactor cooling using condensate '

only. When fission heat is being produced in the- critical core, normal cooling should be provided by forced circulation using either steam drive !

or feedwater-drive of the helium circulators. The current wording of j Specifications 3/4.5.3 and 3/4.7.1.1 do not provide for the OPERABILITY of the feedwater-drive for the safety-related emergency core cooling function ]

when the CALCULATED BULK CORE TEMPERATURE is less than 760 degrees F ]

although fission heat is allowed to be as high as 5% of rated reactor power j in STARTUP. If new specifications are not added to address the OPERABILITY j of the feedwater-drive for the important-to-safety normal cooling function j whenever fission heat is being produced, further restrictions may be 1 I

required.

One option is to delete the footnote on STARTUP in LCOs 3.5.1.1, 3.5.3.1, and 3.7.1.1 and to delete STARTUP in the APPLICABILITY in LCOs 3.5.1.2 and 3.5.3.2. Alternately, to allow the flexibility of performing training starts while limiting the allowed fission heat level, the footnotes on the subject LCOs can be changed as follows. In LCOs 3.5.1.1, 3.5.3.1, and 3.7.1.1, the footnote should be changed to read: "Whenever CALCULATED BULK CORE TEMPERATURE is greater than 760 degrees F, or, in STARTUP, whenever reactor thermal power is equal to or greater than 2 percent of rated reactor power." In LCOs 3.5.1.2 and 3.5.3.2, the footnotes should be changed to read: "Whenever CALCULATED BULK CORE TEMPERATURE is less than or l l equal to 760 degrees F, or, in STARTUP, whenever reactor thermal power is l 1ess than 2 percent of rated reactor power." In addition, the Bases for the affected LCOs should be modified to indicate that the limits on STARTUP are to allow flexibility for achieving criticality during training starts and l that procedures have been implemented to limit the amount of fission heat to much less than 2% during such activity.

4. LCO 3.5.1.2 ACTION b needs to be rewritten as follows:

b. With less than the above required OPERABLE equipment and with no forced circulation being maintained, be in at least SHUTDOWN within 10 minutes and suspend all operations involving CORE ALTERATIONS, control rod movements resulting in positive reactivity changes, or movement of IRRADIATED FUEL, and either:

1. Restore forced circulation on at least one loop prior to reaching a CALCULATED BULK CORE TEMPERATURE of 760 degrees F, and comply with ACTION a, or,

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

l l

1 l

l

a 1 i i

. . 4

5. In LCO 3.5.1.2, both ACTIONS a and b, the words " CALCULATED BULK CORE TEMPERATURE" are used with regard to the required restoration of equipment ]

or conditions. In both instances, a footnote symbol should be added after 1 the word " TEMPERATURE" with the following words provided in the text of the footnote:

Specification 3.0.N provides the methodology and necessary data to q determine the appropriate time interval to reach a CALCULATED BULK CORE TEMPERATURE of 760 degrees F.

PSC needs to provide the appropriate number "N" for the cross-referenced specification.

PROPOSED RESOLUTIONS 1./2./3./4./5. PSC needs to incorporate the required changes.

3. PSC needs to provide additional specifications per the previous cited comment on Section 3/4.4 or to incorporate the suggested changes in the LCOs cited or to propose and justify alternatives. !

NRC COMMENTS-LCO 3.5.3.1 {

1

1. In SR 4.5.3.1, the reference to Specifications 4.5.2.1.s needs to be deleted and replaced with a reference to Specification 4.5.1.1.b.1. I

2. See above NRC Comment No. 3 on LCO 3.5.1.2.

PROPOSED RESOLUTIONS i

1. PSC needs to incorporate the required changes.

2. Same as above for NRC Comment No. 3 on LCO 3.5.1.2.

NRC COMMENTS-LCO 3.5.3.2 l

1. The ACTION for LCO 3.5.3.2 needs to be deleted and replaced as follows:

With both the emergency feedwater and emergency condensate header inoperable, be in at least SHUTDOWN within 10 minutes and restore at least one header to OPERABLE status prior to reaching a CALCULATED BULK CORE TEMPERATURE of 760 degrees F and suspend all operations involving CORE ALTERATIONS, control rod movements resulting in positive reactivity j changes, or movement of IFRADIATED FUEL.

2. In the ACTION, PSC should add the same footnote as described above in NRC Comment No. 5 on LCO 3.5.1.2. Similar information and cross references as provided in the footnote should also be included in the Basis.

3. In SR 4.5.3.2, the reference to Specification 4.5.2.1.a needs to be deleted and replaced with a reference to Specification 4.5.1.1.b.1.

4. See above NRC Comment No. 3 on LCO 3.5.1.2.

l

5 PROPOSED RESOLUTIONS 1./2./3. PSC needs to incorporate the required changes.

4. Same as above for NRC Comment No. 3 on LCO 3.5.1.2.

NRC COMMENT-LCO 3.5.4 In ACTION b (the second part of the ACTION statement on page 3/4 5-30), PSC should add the same footnote as described above in NRC Comment No. 5 on LCO l 3.5.1.2. A request to modify this ACTION has been provided in the NRC Request for Additional Information at Enclosure 2. The words "within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months " need to be replaced with the words " prior to reaching .a CALCULATED-BULK CORE TEMPERATURE of 760 degrees F but within a period of time not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ."

Similar information and cross references as provided in the footnote should also be included in the Basis.

PROPOSED RESOLUTION PSC needs to incorporate the required changes.

NRC COKMENTS-LCO 3.6.2.2

1. In LCO 3.6.2.2, the condition statement should be deleted and replaced with the following:

The Reactor Plant Cooling Water (RPCW)/PCRV Liner Cooling System (LCS) shall be OPERABLE with:

a. One RPCW/PCRV LCS loop OPERATING with at least one heat exchanger and one pump in each loop OPERATING, and

b. With firewater supply available via one OPERABLE flow path.

The change is necessitated to comply with the NRC guidelines for PSC commitments with regard to proposed revisions to existing LCO 4.1.9. These guidelines are given in the NRC letter, Heitner to Williams,- dated December 5, 1986. )

I

2. In LCO 3.6.2.2, the ACTION statement should. be deleted and replaced '

with the following: 1 With no RPCW/PCRV LCS loop OPERATING, within 10 minutes, be in at least i SHUTDOWN and suspend all operations involving CORE ALTERATIONS, control j rod movements resulting in positive reactivity changes, or movement of IRRADIATED FUEL, and:

a. Restore at least one loop to OPERATING status prior to reaching a CALCULATED BULK CORE TEMPERATURE of 760 degrees F, or

b. Restore forced circulation cooling prior to reaching a CALCULATED BULK CORE TEMPERATURE of 760 degrees F.

6 The change is necessitated to comply with the NRC guidelines for PSC commitments with regard to proposed revisions to existing LCO 4.1.9. These guidelines are given in the NRC letter, Heitner to Williams, dated December 5, 1986.

3. If the interfacing isolation valves between the firewater system and l the RPCW/PCRV LCS are not covered in the surveillance on the SAFE SHUTDOWN l COOLING water supply system per SR 4.5.4.1.f or SR 4.5.4.1.g.3, the subject I isolation valtes need to be covered by revising SR 4.6.2.2 appropriately. l l

4. In the ACTION, PSC should add the same footnote as described above in NRC Comment No. 5 on LCO 3.5.1.2. Similar information and cross references I as provided in the footnote should also be included in the Basis. l PROPOSED RESOLUTIONS 1./2./3./4. PSC needs to incorporate the required changes.

NRC COHMENT-LCO 3.7.1.1 See above NRC Comment No. 3 on LCO 3.5.1.2.

1 PROPOSED RESOLUTION l j Same as above for NRC Comment No. 3 on LCO 3.5.1.2.

l NRC COHMENT-LCO 3.7.1.6 l

In the ACTION, PSC should add the same footnote as described above in NRC Comment No. 5 on LCO 3.5.1.2. Similar information and cross references as provided in the footnote should also be included in the Basis. l PROPOSED RESOLUTION 1 l PSC needs to incorporate the required changes.

NRC COMMENT-LCO 3.7.4.2 In the ACTION, PSC should add the same footnote as described above in NRC comment No. 5 on LCO 3.5.1.2. Similar information and cross references as provided in the footnote should also be included in the Basis.

PROPOSED RESOLUTIONS PSC needs to incorporate the required changes.

6, Amendment No. ENCLOSURE'A

,. Page 3/4.5-SAFE SHUTOOWN COOLING SYSTEMS D$lh?Y -

~

3/4.5.1 SAFE SHUTDOWN COOLING EQUIPMENT LIMITING CONDITION FOR OPERATION ,_

3. 5.1.1. a . Two primary coolant loops shall be OPERABLE, each with at least: .

1.

Oneg1tum circulator OPERABLE, and

2. steam generatorsection[( he economizer-evaporizer-superheater (EES) g the reheater)

OPERA 8th pcl "D #'

b. For OPERABLE helius circulators, the following safe I QPERABLE shutdown cooling drives and auxiliary equipment shall be OPERABLE:

flew paths

1) A safe shutdown cooling drive with the capability of providing the equivalent of 8000 rps circulator speed

~

at atmospheric pressure to two circulators e simultaneously, 7O[\ A p Y#

gC Un 2) Jw6 safe shutdown cooling drivef with the capability of providing 3% ra.ted helium flow at operating g *, M q pressure with firewater supply, including two Otd\

OPERABLE emergency water booster pumps and two

' oPEarsLE fiow pathi.

' 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

s Ar.endment No '

Page 3/4.5-DQ;* *~' ~

v

[7 b) At least 10% of primary coolant fee ? -

nT pressure i L

V a boundary bolting and other struc tural bolting which has been removed for the inspection above M { and which is exposed to the primary coolant shall be nondestructively r tested for g identification of inherent or developed defects.

t' q c) Reports kL w

of Within 90 days of examination completion, a N Wh Special Report shall be submitted to the NRC in 4.

accordance with Specification 6.9.2. This

& I report shall include the results of the helium circulator examinations. j i

Nn [f' b.

The steam generators shall be demonstrated ODERABLE:

v j

-e y 1.

1 At .least once per la months be l n, through the emergency feedvater/ verifying proper flow

,f , header and emergency 7 condensate header to the steam generator :ections.

l l

J*#

L

g/. At*1 east once per 5 years by volumetrically examining i *Y$ the accessible port'ons of the following bimetallic i V

93- . - --

M welef

4) t for indications af subsurface defects:

"w p]. .

9) The . main l

steam ring header collector to

,,r 3 , collector drain piping weld for one steam generator module in each loop, and f ..f n{ c. 8) ' The same two steam generator modules shall be re examined at each interval.

v a. v

.r "

b A .f The initial examination shall be performed during SHUTS.

cycle 00WN or REFUELING prior to the beg. inning of fuel This initial examination shall also include the bimetallic welds described above for two additional steam generator modules in each loop.

1 g, /. Tube Leak Examination ,

Ea'ch time a steam generator tube plugged due to a leak, specimens from the accessible connected to the leaking inaccessiblesubheader tubes tubes shall be metallographically examined.

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

MI /eget one d pee 18 menThs by Veelfy 'n 3 tb O PER ABILI.T y o f ea c h supei h eate r o gt h t f lo w p aTh by v e r ! 4 y 'e n 3 .T h o r t h e valver in 't h c. six ischpenT lines can b e o p en ed u d Th aT Th e e M .[. [o vv f3Thf 2PC hot ohyTtuCTod

a,. .- .

Amendment No.

Page 3/4.5-SAFE SHUT 00VN COOLING SYSTEMS

~

3/4.5.1 SAFE SHUTDOWN COOLING EQUIPMENT g3-FEB2 ~ " '

LIMITING CONDIT10N FOR OPERAT10N 3.5.1.2 a. At least one primary coolant loop shall be OPERABLE.-

including at least:

1. One helium circulator OPERABLE, and

2. One steam generator section (e'ner the econostzer-

, g , j f ,,

evaporator-superheater (EES) or reneater) OPERA 8Lp one OPER A3 tg b. For at least one OPERA 8t.E helium circulator, the following k

emergency drives and auxiliary equipment shall be OPERA 8LE:

I .g safe shutdown cooling drive with the capability of providing with 3% rated helium flow at operating pressure firewater supply, including one OPERABLE emergency water booster pump and one OPERABLE flow path, i 2g. The turbine water removal system, including one turtdne i

water removal pump, {

FOR 2 /. The normai bearing water syste.. inciuding oee source of

,% vn bearing water makeup and one bearing water makeup pump (P2105 or P2108), and O ?y,

t ' Yg. The bearing water accumulator (T-2112. T 2113 T-2114 .

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

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

!* - Amend;ent No.

Page 3/4.5-Q.R,W FEB2A *-

ACTION: a. With less than the above required OPERABLE equipment, and y ;ts: , to with forced circulation maintained, be in at least SHUT 00VN e..mJ'eteif, and restore the required equipment to OPERABLE

"'"^ .

status prior to reaching a CALCULATED BULK CORE TEMPERATURE K of 760 degrees F, or suspend all operations involving CCRE ALTERATIONS,. . Conn / t octrwovemerits resulting in positive re' activity changes, or moveunLof JRRA01ATED FUEL.

<N)(. 4

[ t e t) V eq 1 t L eq pm t an ar op n t

~' ' ea t

' ad t s o p t o re h rty A 8 0 f " "

y g ,, g, 1/ /-/

s

/

v v *1 hhs/orVvheMf I IA ,

2.

Initiate PCRV depressurization in accordance with the

'\ '

time specified in Figures 3.5.1-2 or 3.5.1-3, as applicable.

SURVEILLANCE REQUIREMENTS '

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

b.

With less than the above required OPERABLE forced equipment and with no circulation being maintained, be in at least SHUTDOWN within 10 minutes control rodand suspend all operations involving CORE ALTERATIONS, movement of IRRADIATED FUEL, and either: movements resulting in positive re

1. i Restore forced circulation on at least one loop prior to reaching a and comply with ACTION a, orCALCUIATED BULK CORE TEMPERATUR w p 's pc 'C , g ra .< C -

Specification 3.0 rovides the methodology and necessary data to determine the appropriate time interval to reach a CALCULATED BULK CORE TEMPERATURE of 760 degrees F.

I

Amendment No.

Page 3/4.5-Depressurization

min gggg In the unlikely event that all forced circulation is lost for 90 minutes, start of depressurization is initiated as a function of prior power levels, with 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months from 1005 RATED THERKAL POWER being the most Ifmiting case. Operators will continue attempts to restore forced circulation cooling until 5 hours after the loss of. forced circulation.

sources and Multiple flowpaths to estabitsh forced convection cooling using circulators makes required depressurization highly unlikely. Cooldown using forced circulation cooldown is preferred to a depressurized cooldown. with the PCRV liner cooling system. Depressurization of the PCRV under extended loss of forced circulation conditions is accomplished by venting the reactor helium through a train of the helium !

purification system and the reactor building vent stack filters to atmosphere. Start cf depressurization times from ,

l various 3.5.1-1, reactor power conditions are delineated in Figures ;

3.5.1-2, and 3.5.1-3 and are discussed in the FSAR Section 9.4.3.3 and Appendix 0. j Redundancy Criteria The use of 760 degrees F CALCULATED BULK CORE TEMPERATURE as a division between the APPLICABILITY of Specification 3.5.1.1 verses 3:5.1.2 is explained as follows:

l In the FSV HTGR, the limiting parameter of interest is a core inlet temperature greater than 760 degrees F. The CALCULATED BULK CORE TEMPERATURE is a conservative calcu!ation of the maximum potential temperature in the core and surrounding components. The conservatism are such that if the CALCULATED l BULK CORE TEMPERATURE is limited to 760 degrees F, the design inlet temperature of 760 degrees F is not exceeded. Systems l 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 system for F, it is acceptable to require only one OPERABLE accident prevention and mitigation without single failure consideration, cooling requirements. on the basis of the Ifmited core nC%/ P5C ~C o Prov*ddl W

Specification 3.0hprovides the methodology and necessary data to determine the appropriate time interval to reach a CALCULATED BUlX CORE TEMPERATURE of 760 degrees F. If the active core remains below this temperature, which corresponds to the design maximum core inlet temperature as indicated above, then the design core inlet temperature cannot be exceeded and there can be no damage to fuel or PCRV internal components regardless of the amount. including total absence, or reversal, of primary coolant helium flow.

Amendment No.

. Page 3/a.5-bb,0..UY i

All IEB 2 3 '??5 forced circulation may be interrupted for maintenance .

p trposes provided that the time calculated for CALCULATED BULK CO?E TEMPERATURE to reach 760 degrees F is not exceeded.

Howtver, if forced circulation. is temporarily res tored, a recalculation shall be performed, based on present conditions, to ettablish a new time period for CALCULATED SULK CORE TEMPE.1ATURE to reach 760 degrees F.

also l'e taken out of service for maintenanceRedundant systems may testing provided or. surveillance that forced circulation is maintained. The time t > reach CALCULATED BULK CORE TEMPERATURE to 760 equal degreer F may be recalculated as often as required. :

Steam Gt nerators DeleT t ad Th st I g er rehe f $ c., n ee)s Reptace om the the s cia ed EES gen cti ns an ec ve t em e en ea r w t k ., per '

ater ead wh are r ir be T. p r w., J e pec ' at n. stem n ,,

b v ify (1 P

ITY s te e#

fr ch of t for e on d r gy 0

ead st o h ch am 9 ra r ?"8F2fb Bimetallip Weld Examination #

en < i t lu k ved liaet The steam generator twJ crossover tube bimetallic welds between SqperTIS Incoloy 800 ud 21/4 Cr-1 No materials are not accessible for examination.

rin The bimetallic welds between steam generator header collector, the main steam piping, and the

, g' t1 col ector drain piping are accessible, 2 naty rit materials, at d operate at conditions not involve the same different fria the crossover tube significantly collector drain piping weld is also geometricallybimetallic similar to welds. The the crossover tube weld. Although expected defects te occur, this specification allows for detectionminimal wafch of degradation affect b; metallic might result from conditions that can uniquely welds made between these AdditioniI collector welds are inspected at thematerials. initial examination to establish should defects be a baseline which could be used, examinations subsequently be required.found in later inspections and add Whenever the CALCULATED BULK CORE TEMPERATURE both the reheater and EES sections of theexceeds 760 degrees F, steam generator must be OPERABLE. The steam generator reheater water from either the emergency condensate or EES sections can receive i feedwater header as required to be OPERABLE per thisheader or the emergency per Specification 3/4.5.3. System Specification and verifying flow from each of the aforementionedflow OPERABILITY is determined by LCO 3.5.3.1) through each section of each steamemergency generator headers (see BULK CORE TEMPERATURE is lessWhenever orthe CALCULATED or the than rees equal F to 760 deg i determined headers (see LCO by verifying 3.5.3.2) flow from either of the gency aforemention generator. through either section of either steam j

m .. . . . . _

s . c ; .: .

- Page 3/4 5-26

3.

DRAFT SAFE SHUTDOWN COOLING SYSTEMS 3/4.5.3 EMERGENCY CCNCEN$ ATE AND EMERGENCY FEE 0 WATER HEAC !

EMERGENCY CONCENSAIE AND EMEAGENCY FEECWATER WEACE45 -- ---- . !~urt L!4tTING CCN0ir:04 F'9 C:E4aTICN 3.5.3.1 The emergency concensate header anc the emergency feedwater header shall be CPERABLE.

APPLICABILITY POWER, LO 00WER, STARTUP", and SHUT 00VN*

ACTION:

With either the emergency condensate header or the emergency feedwater header inoperable, restore the Inoperable header to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or:

1.

When in POWER, LOW POWER, or STARTUP, be in at least SHUT 00WN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or

2. When in SHUT 00WN. suspend all ope ra tion s involving control rod movement resulting in positive reactivity changes, or movement of IRRADIATED FUEL.

/ .

\*'

SURV$ILLANCE REQUIREMENTS-4.5.3.1 No accitional Surveillance Requirements are required other tnan those surveillance identified in Specification t.5.2.;'

~**-

9, 5, l , I, b . I

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

b F

r~t- '2.

- Page 3/4 S-27 -

1 DRAFT SAFE SHUTOCWN COOLING SYSTEuS M3 0 2 3/4.5.3 EMERGENCY CONDENSATE AND EMERGENCY FEE 0 VATER HEADERS EMERGENCY CONDENSATE AND Eu! AGENCY FEE 0 WATER HEADERS . - SEMp

-. i LIMITING CCN0! TION FOR CPERAT!CN 1

3.5.3.2 Either the condensate header or the emergency feedwater i header shall be OPERABLE. i I

APPLICABILITY: STARTUP", SHUT 00WN", and REFUELING -

ACTION: With ot, th .er ncy f dwat and gency h der opera e, estor a densa e st one a r 0 LE sta s e to 'im calc ed fo the c 1 fromx ecay he to to he t {

each ng CULATED LK COR '

TEMP AU 760 deg F an spen all e ations i

in tvi' CC t A ". TIONS o ent I ro vement es ting ce iti reac 1 EL ty chang s, or move ent or RRADIATED' 1

i SURVEILLANCE REQUIREMENTS s ...;! !

4.5.3.2 No additional Surveillance Requirements are required other than those surveillance identifiec in Specification 1 t ? i : .%

q. f.l . l . b. l "Whenever 760 degrees F.CALCULATED BULK CORE TEMPERATURE is less than or eRual p 3 C n c.ed s T. p r o v '. J c. T.

/

Specification 3.0hprovides the methodology and necessary data to determine the appropriate time interval to reach a CAlfU1ATED BUlX CORE TEMPERATURE of 760 degrees F.

t.

I

~

y _. -

{ Vith both the emergency feedwater and emergency condensate header inoperable, be in at least SHUTDOWN within 10 minutes and restore at least one header to OPERABLE status prior to reaching a CAlfUIATED BUIX CORE TEMPERATURE @of 760 degrees F and suspend all operations involving CORE ALTERATIONS, control rod movements resulting in positive reactivity changes, or movement of IRRADIATED FUEL.

-m e r. : e a. : no,

' Page 3/4 5-28 -

DRAFT BASIS FOR SPECIFICATION LCO 3.5.3 / SR 4.5.3 NOV 3 0 W The OPERABILITY of the emergency condensate header and the emergency feedwater header ensures recuecant water supply paths to the helium circulators and steam generators for SAFE SHUT 00WN COOLING of the 'p l a n t . **

In th e" e v e n t"*Tf"-*a failure of the normal. feedwater lisfe, the availability of-either the emergency feedwater or e.eargency condensate lines provides aceovate shutcown capaollity. OPERABILITY of the i aforementioned verifying flow headers is accomplished caring SHUT 00WN by through each header to the steam generators and helium circulators.

(In the FSV HTGR, the limiting parameter of interest is a f core inlet temperature greater than 760 degrees CALCULATED F. The BULK CORE TEMPERATURE is a conservative calculation of the maximum potential temperature in the core and surrounding components.

The conservatism are such that if the degrees CALCULATED BULK CORE TEMPERATURE is limited to 760 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 i

P5C criterionthen greater whenever CALCULATED 760 degrees F. However, BULK CORE TEMPERATURE is i when CALCULATED BULK heeds CORE TEMPERATURE is equal to or less than 760 degrees F, it is acceptable to require only one OPERABLE system for Io accident prevention and mitigation consideration, on the basis of wi. sut single failure c r o s s .: requirements, tne 11afted core cooling r e f e r ea<

All fo'rc ed circulation may be inter .:ted for maintenance

! SP " h U" purposes provided that the time calculated for CALCULATED BULK CORE TEMPERATURE to reach 760 degrees F is not i

?* O, d exceeded. However, if forced circulation is temporarily restored, a recalculation shall be performed, based on present conditions, to establish a new time period for CALCULATED BULK CORE TEMPERATURE to -each 760 degrees F.

Redundant systems may also be taken out of service for maintenance or surveillance testing p*evided that forced circulation is maintained; and the time to reach CALCULATED BULK CORE TEMPERATURE equal

( recalculated as often as required. to 760 degrees F may be The emergency feedwater header is not normally placed in service until approximately 30% reactor oewer, to prevent unnecessary long-term wear of components associated with the emergency feedwater header.

Nevertheless it is still required to be OPERABLE during the aforementioned L400ES.

1

. -... . . ... e t N :

4 page 3/

NOV 3 0 285 With degrees F:CALCULATED BULK CORE TEMPERATURE less than or equal'to 760

4. With one of the above required pumps and/or makeup conds -

Inoperacle, restore the inoperable e.:utpfe'nt to l status witn CPERABLE 14 days or provide an alternate backup pumo or water suoply. The provisions of Speci f f cation !

are not apolicac'e.

3.0,4 !

b. 1 With the SAFE SHUTCOWN COOLING water system otherwise inoperacle, establish backup i a system for fire suppression purposes U +hia N % s *- ]

)

SURVEft.LAN REQU!REMENTS );

4.5.4.1 The SAFE SHUT 00WN COOLING water supply system shall be demonstrated OPERABLE:

a.

At least once per 7 days by verifying the contained water

- supply volume in each of the circulating water makeup ponds, b.

At least once per 31 days by starting the electric motor-driven fire pump and operating it for at least 15 minutes.

c. At least once per 31 days by starting each circulating water makeup pump that is not alrea:/ running, d.

At least once per 31 days by verify %g that each valve in the flow path, that is not locked, sealed, or otherwise i secured in place is in its correct position, !

l

e. At least once per 12 months by performance of a system flush, f.

At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel, g.

At least once per 18 months by performing a system functional I Actuation test which includes simulated automatic of the system throughout its operating sequence, and: I 1.

Verifying that the automatic valve in the flow cath actuates to its correct position, _

2.

Verifying that each pump (motor-drAi/en and engine-driven) develops at least 1a25 gpm at 119 psig, prior to reaching a CA14UIATED BULK CORE TEMPERATUREkf 760 de6rees but F within a period of time not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

g P5c w e ,4 s T , P r. v t / c d",

Specification 3.0 provides the methodology and necessary data to determine the appropriate time interval to reach a CALCULATED BULK CORE TEMPERATURE of 760 degrees F.

~

.. Y

. NOV 3 0 tan i With two out of three circulating water makeup pumps inoperableI or with any one firewater circulating water makeup pond inoperable pump or a inoperable and the C4LCULATED BULK CORE

TEMPERATURE greater than 760 degrees F, a restoration .

72 Dumos.

hours is considered sufficient based on redundant timeflow of paths ,

and i However, with all circulating water makeup pumps, beacers, or firewater pumps inoperacle, a restoration time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months sucply is specified, are lost. as all means of SAFE SHUT 00WN COOLING water The surveillance that all equipment, water supplies, and flow paths willidentifiei OPERABLE as specifled in order remain }

COOLING requirements specified above.to meet those $AFE SHUT 00VN I

The use of 760 degrees F CALCULATED BULK CORE TEMPERATURE as a division between the ACTIONS is explained as follows:

In the FSV HTGR, the limiting parameter of interest is a core inlet temperature greater than 760 degrees F. 3 BULK CORE TEMPERATURE is a The CALCULATED I maximum potential temperature conservative calculation of the P5C components. in the core and surrounding i bce45 :

EULK CORE TEMPERATURE is limited inlet temperature of to 760 degrees F the design .

7' used for accident prevention and760 degrees F is not exceeded. Systems mitigation are required to

k. )' satisfy the l c r o is -

/

CORE TEMPERATURE is greater thanHowever, 760 degrees'F. when single i

tefere*( c CALCULATED BULK degrees F, it is acceptable to require onlCORE TEMPERATURE is for accident y one OPERABLE system gf .4. g consideration, prevention and mitigation without single failure requirements. on the basis of the limited core cooling j j

2**

All forced circulation may be interrupted for maintenance

): purposes TEMPERATURE provided that the time calculated for CALCUl.ATED BULK l to reach 760 degrees F, is not exceeded. However, i if forced circulation is temporarily restored. 4 recalculation {

f snall be performed, based on present concisions, to establish a i new time period for CALCULATED BULK CORE TEMPERATURE 760 degrees {

.i F. to reach j Redundant systems may also be taken out of service for maintenance or surveillance testing provided that i forced circulation is maintained; !

and the time to reach (CALCULATEDBULKCORETEMPERATUREequalto760degreesFm recalculated as often as required. be l

i I

i i

i 1

I

, , _ ~ _ - - - -

Amendment No.

Page 3/4 6- I DRAvr

\

PCRV AND CONFINEMENT SYSTEMS FEB 2 e p"*c 3/4.6.2 REACTOR PLANT COOLING WATER /PCRV LINER COOLING S SHUTOOWN t.IMITING CONDITIONS FOR OPERATIONS -

3.6.2.2 I

g APPLICABILITY: STARTUP SNUTDOWN[*and REFUELINf (

ACTION: it o RPCW C i P I s t 1

pr o T es or at ta a in a o IM5tr CO P E O e p 1 er O g I iv vin 0 E I 5 a ~ - r n o ve ea t ity c ng g. "/ \ ' / t g SURVE!LLANCE REQUIREMENTS- -

7 4.6.2.2 No additional surveillance requirements other ths.n those identified per Specification 4.6.2.1.

t a

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

[ Th y c es e ut h e on s

f in in ys .ay s a E a o E

" s s n e a o P5 n e.< 4 : T

P r*

d < "N ?

M Specification 3.0 de te rtaine rovides the methodology and necessary data to the appropriate time interval to reach a CALCULATED BUlX CORE TEMPERATURE of 760 degrees F.

tVM AFO Ois _Y

l L c o 3. 4. 2. L Ins erT A The Reactor Plant Cooling Water'(RPCW)/PCRV Liner Cooling System (14S) shall be OPERABLE with:

l a. One RPCW/PCRV LCS loop OPERATING with at least one heat exchanger- ,

and one pump in each loop OPERATING, and

'1

b. With firewater supply available via one OPERABLE flow path. j I n ser r G J With no RPCW/PCRV LCS loop OPERATING, within /0 minutes, be in ' at least SHUTDOWN and suspend all operations involving CORE ALTERATIONS, control rod movements resulting in positive reactivity changes, or movement of 1 IRRADIATED FUEL, and:

J

a. a RestoreatleastonelooptoOPEy1 TING status prior to reaching CALCULATED BULK CORE TEMPERATURE of 760 degrees F, or

b. tion cooling prior to reaching a CALCULATED Restoreforcedcirculgof760degreesF..

BUlX CORE TEMPERATURE i

1 ...~- .

Amendment W .

Page 3/4 6-DRAFT

[The use of TCn s e f division between the APPLICABILITY of Specification 3.6.2.1 760 degree 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 f5g degrees F. The CALCULATED a BULK CORE maximum TEMPERATURE potential temperature in the is a conservative calculation of the hcddy components. core and surrounding The conservatism are such that if the CALCULATED To i

BULK CORE TEMPERATURE f s limited to 760 degrees F, the design inlet temperature of 760 degrees F ts not exceeded. Systems used U"U' for accident prevention and mitigation are required to satisfy g efer.nct the single failure criterion whenever CALCULATED 80LK CORE TEMPERATURE is greater than 760 degrees F. However, when' CALCULATED BULK CORE TEMPERATURE is equal to or less than 760 gfCg' "g,, degrees F, it is acceptable to require only one OPERABLE system 3O'p* required for accident prevention and mitigation as acceptable without single core cooling failure c nsideration, on the basis of the limited 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, tp establish a new tire period CORE TEMPERATURE for CALCULATED BULK also be taken out of serviceto reach 760 degrees F. ~ Redundant systems may for maintenance or testing provided that forced circulation is maintained. surveillanceThe time to reach CALCULATED BULK CORE TEMPERATURE equal to 760 F degrees may be recalculated as often as required.

I I

l

t. ;.

Amendment No.

Page 3/4.7-DRAFT PLANT AND SAFE SHUT 00WN COOLING SUPPORT IE3 2 8SYSTEMS G33 3/4.7.1 TURBINE CYCLE SAFETY VALVES - SHUT 00WN 4

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: SHUT 00WN and REFUELING ACTION: With less than the above required safety vgive(s) OPERABLE, restore the required safety valve (s) to OPERAB1J status prior F or suspend all operationsto reaching a CALCULATED BUL involving CORE ALTERATIONS or con trol rod movements resulting in positive reactivity changes.

SURVE!LLANCE REQUIREMENTS- -..

4.7.1.6 No additional per Speci surveillance fication 4.7.1.5. required beyond those identified M lPSCnecdr T

p r o v

d e " [.'

Specification 3.0@ provides the methodology and necessary data to determine the appropriate time interval to reach a CALCULATED BULK CORE TEMPERATURE of 760 degrees F.

COR v

INFO ONLY l

s 1

1

Amendment No.

Page 3/4.7-

' MAFT 1 BASIS FOR FEB 3 81980 SPECIFICATIONS 3.7.1.6/$R a.7.1.6 LCO 3.7.1.5/SR 4.7.1.5 AND LCO The economizer-evaporator-superheater (EES) section of eacn steam generator loop is protected by three spring-loaded safety valves, each with one-third nominal relieving capacity of each loop. The j reheater section of each steam generator loop is protected from i ove rpressure transients by a single safety valve. These steam (

generator 10.2.5.3. safety valves are described in the FSAR, section :

l The above valves are required to be tested in accordance with (ASME Section XI, maintenance. IGV requirements) every 5 years or after tested with steam.

To satisfy the testing criteria, the valves must be j in Since these valves are permanently installed i steas piping, the appropriate means for testing require plant i 1

, power to be in excess of 22% RATED THERMAL POWER. Thus, the test-must be conducted during LOW POWER. Conditions are specified so as to minimize operation at power untti the valves are tested.

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

Ouring 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 ove' pressurization of any steam generator section (i.e., one boiler feed pump per operating loop). Conversely, with any reheater safety valve inocerable, plant operation is restricted

. to a more restrictive Mode. .

1 A 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action time for repair or SHUTDOWN due to inoperable safety in valves ensures that .these valyes are returned to service a relatively short period of time, during which an 1 overpressure hours does nottransient is unlikely. Operation at pewer for 72 for any extended period. result in a significant loss of safety function The setpoints for the safety valves identified in Table 4.7.1-1 are those valves identified in the FSAR with tolerances applied such that the Technical Specifications incorporate an upper bound setpoint. This is consistent with Et operating limits in these Specifications., incorporating normal

, zo yfovide t 1T o * * \ ' S**

C ALC UL AT E D B U L le. c. o R E TEMPER-USin3 .

ATvRE- aw2 Io cro rr -r ef eren ce 4 pee 14 ; e n T io " 7 . 0, N ,

1 l

,,_-_- Q

e...c nc.ter.?, NC .

,0+

Page 3/4 7-21 PLANT AND SAFE SHUTOCVN COOLING SUPPORT SYSTEMS 3/a.7.4 SERVICE WATER SYSTEM-SHUTOCVN NOV 3 0198'j LIMITING CONDITION POR OPERATICN -

3.7.4.2 The service water system shall be OPERABLE with:

a. One service water pump (P-4201, P-4202, or P-42025)

OPERABLE, b.

An- OPERABLE flow path to SAFE SHUTOOWN COOLING users of service water emergency diesel instrument coolers, air compres(sor and after coolers, and the reactor plant cooling water /PCRV liner cooling heat exchangers; E-4601, E-4602, E-4603, and E-4604), and

c. .

An OPERABLE flow path from the circulating water makeup system to the service water pump pit.

APPLICABILITY: STARTUP*, SHUTOOWN", and REFUELING i

ACTION:' With any i of the requirements identified in 3.7.4.2 a, b, (c . - or c above inoperable, restore the e uipment to OPERABLE

~

status prjpr to reaching TEMPERATURE"of 760 degrees F, andasusoend CALCULATED BULK CORE all operations involving CORE ALTERATIONS or control rod movements resulting in positive reactivity changes.

'Vith theF.CALCULATED BULK CORE TEMPERATURE less than or equal to 760 degrees pc y ,,., n y s, f Specification 3.0 determine provides the methodology and necessary data to TEMPERATURE of 760 degrees F,the appropriate time interval to reach a 9

+

O g - :ccment No.

Page 3/4 7 24 HOV 3 0 bus i

1 During POWER, BULK CORE TEMPERATURE greater thanLOW POWER, STARTUP, and SHU 760 degrees F, with only enj service water pump OPERABLE, .a restoration time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is provided to restore another pump to OPERABLE status. During this 72 l I

hours, service water needs can be met by the redundant pump.

With an inoperable flow path to the emer instrument air compressors and after coolers, gency diesel coolers or cooling is required, initiation of backuo because tne affected component may be automatically initiated tp perform SAFE SHUT 00WN COOLING.

inoperable flow path With an cooling so that heat exchangers, backup cooling capability must beto the reactor verified, it can be manually initiated if regured. Actual initiation

! of backup cooling to the PCRV LCS heat exchangers is undesirable except fouling. inSeventy-two an actual emergency to minimite possibility of tube OPERABLE hours is provided to restore the flow path to status, this provided for in the system OPERABILITY is consistent with the restoration times components. If the flow requirements for the above backup cooling will be initiated path cannot be restored within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , verified within (firewater) or the capability Both firewater pumps must be OPERABLE.1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to restore the equipment ;

(

for The Surveillance Requirements '

firewater pump OPERABILITY are given in Specification 3.5.4 neither of these conditions can be met, the plant must be in SHUT 00VN If within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If only one firewater pump is OPERABLE, a second pump must be restored to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or the plant must be in SHUT 00VN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

/' The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restoration time is required because service water is un av a.il abi e , and the firewater supply has no redundant capability. If no firewater pumps are OPERABLE, the plant must be in SHUT 00VN witnin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months because service water and backup cooling (firewater) are both unavailable. i With the flow servicetowater provided restorepump pit inoperable, a restoration time of 24 hou is * '

the flow path to CPERABLE status.

( time is adequate, since backup cooling (firewater) can be This ACTION and considering the makeup requirements of the service water system. initiated, w e<M [0uring STARTUP, SHUTOOWN, and 7' CORE pump, TEMPERATURE 1ess than or equal to 760 degrees F, a ser crus- a flow path from the circulating water makevo system toflow the service water pump pit Wg,,,qf calculated toare reachrecuired to be OPERABLE prior a CALCULATED BULK to theCORE time TEMPER F. of 760 degrees ;

Y , eg ,. g ;,, Thiswhen ACTION ensures that the plant will remain in a stable condition service 3 , 9, bl , testing, or unanticipatedwater is unavailable due to maintenance, outages.

j No specific surveillance are required as the service water' system is normally flow pathoperating, including the reautred OPERABLE flow paths.The by the surveillance testing of tne diesel generators.to the stancey d L-________----_-_-_-__-______ _ _ _ _ _ _ _ . _.

{

l ENCLOSURE 5 j 1

l Categorization of the Auxiliary Electrical Power Systems, 3/4.8, j comrnents as transmitted to PSC by the NRC Letter of May 6,1987, i

Enclosures 1 and 2. Categorization symbols are the same as in previous l meetings and correspondence (for example, the C. Hinson memorandum of December 15, 1986). l l

1

'1 1

1 1

1 I

l l

i 1

1 1

]

1 l

Conrnent No. _

Categorization ENCLOSURE 1 1 D 2 0 3 0 4 D l 5 0 6 0 1 7 0 j ENCLOSURE 2 1 B 2 D l 3 0 I 4a B )

4b B l 4c D l 2

4d B 4e B 5 0 6 D 7 D*

8 D l

9 D 10 0 1 11 D l l 12 D l 13 D I 14a f I 14b,c D 15 D j 16 0 for reactivity -

, change portion 17 D 18 0 19 0 20 0 21 0 22 0 23 0 24 D 25 0 26 0 27 D 28a D*

28b B 28c D*

2

1 e

r' 4

Connent No. Categorization ENCLOSURE 2 (Continued) 1 28d 6 28e B 28f A# DA# for Licensee action to address 5 days vs 2 hrs. !.

l 28g 0 29h B

29) AI 29j B l

l l

)

i i

l i

)

1

-l,

)

3 I

I J

o UNITED STATES

+f [og 1

- ! g NUCLEAR REGULATORY COMMISSION

W ASHINGTON, D. C. 20555 p

,/

July 2,1987 Docket No. 50-267 Mr. R. O. Williams, Jr. ;

Vice President, Nuclear Operations l Public Service Company of Colorado l Post Office Box 840 i i

Denver, Colorado 80201-0840 l l

Dear Mr. Williams:

1

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR TECHNICAL SPECIFICATION UPGRADE PROGRAM FOR FORT ST. VRAIN NUCLEAR GENERATING STATION j

(TAC NO. 56565) 1

References:

See Enclosure 1 J We have reviewed the information you submitted with your letter, Reference 1, on the Technical Specification Upgrade Program for the Fort St. Vrain Nuclear-

)

1 Generating Station (FSV). Your submittal, provided documentation of comment )

resolutions that were reached between Public Service Company of Colorado (PSC) i and the Nuclear Regulatory Commission (NRC) during a meeting on October 27-30, 1986, regarding the FSV Technical Specification Upgrade Program (TSUP). This ,

I meeting was held to resolve comments provided by the NRC in Reference 2, on the i November 30, 1985 FSV TSUP draft submittal. The NRC provided a meeting summary I in R'.ference 3, which tabulates the comments discussed and the nature of their resolution.

This letter provides a request for additional information (See Enclosure 2) as a result of our review of your current submittal, aga. inst our original comments (Reference 2). The comments are identified in the same manner as used in our l meeting summary (Reference 3). Also, in Enclosure 2 we have provided requested

! actions for PSC resulting from our review and resolution of.the eleven NRC ,

) action items from the meeting of October 27-30, 1986, and for the action items l summarized in Enclosure 1 to Reference 2 that still require further PSC action. 1 Further, Enclosures 3 and 4 contain supplements to our comments transmitted in 3 Reference 4 on the safety-related cooling functions. These supplemental comments address our view of how your proposed revision to Technical Specifica-tion LCOs 4.3.1 (Reference 6) and 4.1.9 (Reference 7) impacts the review of the final draft of the FSV TSUP. This is in response to your desire to have the l

reviewers of the TSUP final draft satisfied that your proposed LCO 4.1.9 resolves any previous comments or concerns raised during the TSUP final draft review. Enclosure 3 provides the supplemental concerns and their proposed resolutions. Enclosure 4 is our markup of the November 30, 1985 draft Technical Specifications that incorporate our proposed resolutions.

-" .Y

i </

=

,e_ A**

1

, , i i

Enclosure 5 transmits our proposed categorization of our comments on the Auxiliary Electrical Power Systems, Section 3/4.8, as transmitted to you in /

Reference 5. This proposed categorization uses the same designations as used !

in previous correspondence, for example, Reference 3. This proposed categoriza- I' tion is for your information and use in preparing your response to these comments. .

Also, PSC should consider transmitting in TSUP format all Fort St. Vrain J Technical Specification Amendments that have been issued in parallel with, but not included in the TSUP final draft of November 30, 1985, and any that may be .

issued in the future, but prior to the TSUP amendment application. As a l minimum this would include Amendment Nos. 47 through 55. Receipt of the j material in the TSUP format will expedite our review and transmittal of any comments prior to this material being submitted in the TSUP amendment application.

Finally, our review of your TSUP amendment application will be expedited further if you would transmit your proposed resolutions to the forty-two "A#"

(PSC action items) and twenty-four "C" (PSC to explain in proposed safety evaluation) Category items (See the list of these items in the October 27-30, j 1986 meeting summary, Reference 4). Again, this will allow us the opportunity to review and comment on these resolutions prior to their being submitted in ,

the TSUP amendment application. j Please provide the required information within 60 days of receipt of this letter. If you feel that further discussion would be helpful in resolving these open issues, please call me at (301) 492-7592.

Tne information requested in this letter affects fewer than 10 respondents; therefore, OMB clearance is not required under P.L.96-511.

1 Sincerely, i l5 i !

Kenneth L. Heitner, Project Manager I Project Directorate - IV Division of Reactor Projects - III, j IV, V and Special Projects Office of Nuclear Reactor Regulation

Enclosures:

As stated 4

cc w/ enclosures:

See next page DISTRIBUTION Docket File gPDR Local PDR E . tLuv PD4 Reading D. Crutchfield F. Schroeder J. CalvoP. Noonan K. Heitner OGC-Bethesda E. Jordan J. Partlow ACRS (10)

J. Miller , R. Emch PD4P;antFile 09 PD4/LA hN PD4/PM h PD4/D 'l TSB PNoonan KHeitner: sr JCalvo ler REmc 7/l /87 7/I/87 7/ v/87 /.1/87 7/J.-/87

9

. Mr. R. O. Williams Public Service Company of Colorado Fort St. Vrain '

CC:

Mr. D. W. Warembourg, Manager Albert J. Hazle, 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 GA Technologies, Inc. Nuclear Production Division Post Office Box 85608 Public Service Company of Colorado l 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 Public Service Company of Colorado Denver, Colorado 80201 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Senior Resident Inspector U.S. Nuclear Regulatory Commission Mr. R. F. Walker P. 0. Box 840 Public Service Company of Colorado Platteville, Colorado 80651 Post Office Box 840 Denver, Colorado 80201-0840 Kelley, Stansfield & 0'Donnell Public Service Company Building Commitment Control Program Room 900 Coordinator 550 15th Street Public Service Company of Colorado Denver, Colorado 80202 2420 W. 26th Ave. Suite 100-D Denver, Colorado 80211 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 )

Chairman, Board of County Commissioners of Weld County, Colorado Greeley, Colorado 80631 1

Regional Representative Radiation Programs Environmental Protection Agency 1 Denver Place 999 18th Street, Suite 1300 .

Denver, Colorado 80202-2413 i i

l ENCLOSURE 1 REFERENCES

1. H. L. Brey letter to H. N. Berkow, Technical Specification Upgrade Program, February 20, 1987, Public Service Company of Colorado, (P-87063).

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

3. Memorandum for H. N. Berkow from C. S. Hinson, Summary of October 27-30, 1986 Meeting at Fort St. Vrain (FSV) to Discuss Staff Comments on the FSV j Technical Specification Upgrade Program (TSUP), December 15, 1986, U.S.

Nuclear Regulatory Commission.

4. K. L. Heitner letter to R. O. Williams, Jr. , NRC Comments on the Technical i Specification Upgrade Program (TSUP), LCOs for Safety-Related Cooling Function, April 17, 1987, U.S. Nuclear Regulatory Commission.

5. K. L. Heitner letter to R. O. Williams, Jr. , Draft Updated Technical Specification 3/4.8, Request for Additional Information (TAC No. 56565),

May 6, 1987, U.S. Nuclear Regulatory Commission.

6. R. O. Williams letter to H. N. Berkow, Proposed Technical Specification Change Eliminating Reliance on Reheater Section of Steam Generator for Safe Shutdown Cooling, January 15, 1987, Public Service Company of l Colorado,(P-87002). j

7. Draft R. O. Williams letter to H. N. Berkow, Proposed Technical Specifi-cation Change, LCO 4.1.9, Core Inlet Orifice Valves / Minimum Helium Flow ,

and Maximum Core Region Temperature Rise, April 17, 1987, Public Service j Company of Colorado, (P-87124). '

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t e a ENCLOSURE 2 ADDITIONAL INFORMATION NEEDED TO COMPLETE REVIEW Of TECHNICAL SPECIFICATION UPGRADE PROGRAM FOR FORT ST. VRAIN NUCLEAR GENERATING STATION The f ollowing is a list of requests for additional information needed to complete this review. The identification of each item is the same as in the NRC notes of the October 27-30, 1986 meeting (G. L. Plumlee Memorandum Dated, December 1, 1986) unless noted otherwise.

TABLE 1.0-1 1 l

The Licensee should provide additional justification for their proposed mark-up of Table 1.1 (P. 1-9, Attachment 1 of PSC letter of February 20, 1987) or should make the "@" and "#" footnotes specific to the precise evolutions (tests) or surveillance involved. PSC references such provisions in the GE-BWR STS. However, neither the GE-BWR STS, NUREG-0123 Rev. 3, nor the Perry TS (example given out in October 27-30, 1986 meeting by PSC) allow indiscriminate switching as would be allowed with the PSC proposed words of "...for the purpose of performing surveillance or other tests...". Both the GE-BWR STS and the Perry TS allow switching of the Mode Switch Position only under rather precisely defined conditions, such as "...to test the switch interlock functions..." or "...while a single control rod drive is being removed from the reactor pressure vessel per i Specification 3.9.10.1. . ." or ".. .while a single control rod is being ;

recoupled or withdrawn provided that the one-rod-out interlock is OPERABLE".

Further, please revise the pound sign (#) footnote in Table 1.1 by deleting the words "provided that k gg is verified less than 0.99" and replacing the deleted words with "provided that the control rods are verified to rena m fully inserted." As acknowledged in the PSC response to NRC Comment No. 4 on TSUP draft LC0 3.1.3, the use of the word " verified" is incorrect in estabitshing a quantitative estimate of SHUTDOWN MARGIN (that is, k,gg) because the reactor operators cannot actually verify the accuracy of the calculated assessments provided to them. However, l

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' consistent with the equivalent footnote provided in Table 1.2 of the BWR-STS (NUREG-0123 Rev. 3), the operators can verif y that control rods are fully inserted.

SL 2.1.1-5 The Licensee should provide additional information and a safety evaluation to support splitting the existing FSV Safety Limit 3.1, part into the proposed Safety Limit Section 2.1.1 (November 1985 Oraft) and part into a Limiting Condition for Operation LC0 3.2.6 (November 1985 Draft).

The 24 hr action time has been adequaiely addressed in PSC's letter of F ebruar y 20, 1987. However, the reference to FSAR Revision 4.Section 3.6.8 does not provide specific justification for downgrading part of the existing Safety Limit to a Limiting Condition for Operation. In fact, FSAR Revision 4 Section 3.6.8 titled " Core Safety Limit" discusses all of the limits in the proposed SL 2.1.1 and LCO 3.2.6 as if they were safety limits )

as does the existing FSV SL 3.1.

LC0 3.0 - New Item Please provide an equivalent LCO to the recently proposed LCD 4.0.4 in tho existing Technical Specifications.

LC0 3.1.1-1 l

Please revise the wording of TSUP draf t LCO 3.1.1.a to add the words j "from the fully withdrawn position" immediately after the words "152 seconds." PSC was to propose an alternative to the May 30,1986, NRC markup of the TSUP draf t. The proposed revisions to the basis does not adequately clarify the limiting condition for operation. In addition, the wording of the proposed insert at the bottom of page 3/4 1-5 in the PSC markup (Attachment 1 to P-87063) needs to be revised to read as follows:

The full insertion scram time can be determined either directly from a full insertion scram time or indirectly from a partial scram time of 10 inches or more. For the partial scram time, the estimate of an 2

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extrapolated scram time is always based on assuming scram from the fully withdrawn position and not from the actual rod position.

LC0 3.1.3-4,-8 l Please provide additional information with regard to the PSC position on the substantiation and verification of nuclear methods as expressed in the PSC response to NRC Comment Nos. 4 and 8 that are given in Attachment 2 to P-87063. Is it PSC's position that the verification and documentation of the assessment methodology used in LCO 3.1.3 neet the intent, if not all 1 the specific requirements or guidelines, of industry standards per ANSI /ASME N45.2.11, ANSI /ANS-19.3 (Section 6), ANSI /ANS-19.4, ANSI /ANS-19.5, and ANSI /ANS-19.6.1? Are detailed records or other supporting documentation for the PSC position maintained from the reviews of the base reactivity curve as required to be performed by the Nuclear Facility Safety Committee (NFSC) per the existing Technical Specifications LC0 4.1.8 and SR 5.1.4? Can it be demonstrated from the NFSC records that the type of analytical-to-experimental comparisons cited in the PSC j position have been and are being factored into the NFSC review? If not, j l explain how the NFSC has verified the continued use of the base reactivity l curve to assure an adequate SHUTDOWN MARGIN per the basis of existing l LCO 4.1.8 without documenting and updating such comparisons as those cited in the PSC position? If such documentation is not being maintained, proposed TSUP draft Specification AC 6.5.2.7.1 needs to be changed such that the NFSC is specifically charged with reviewing and documenting the results of the perf ormance of SR 4.1.3, SR 4.1.4.1.2, SR 4.1.5, and SR 4.1.7 and the results of comparing the reload design against the requirements specified in DESIGN FEATURE 5.3.4. The reviews and documentation are both necessary and appropriate so that there is a record l of the verification process for assuring the adequacy of the SHUTDOWN KARGIN assessment methodology employed in LC0 3.1.3 and SR 4.1.3. This i

l approach is consistent with the intended role of the NFSC as agreed upon l with NRC and acknowledged by PSC in the Attachment to PSC letter, C. K. Millen to R. S. Boyd, September 11, 1975.

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. t LC0 3.1.6 New Item (Old LC0 3.9.1-8)

The Licensee should add SHUTOOWN and REFUELING to the Applicability statement of LC03.l .6 on the Reserve Shutdown System. Such added l applicability should account for the exception required for any two control f rod pairs which may be removed f rom the PCRV (LCO 3.1.4.2.a.1 of Novenber 30, 1985 Draft). Although the May 30, 1986 NRC comment LCO 3.9.1-8 on inadequate SHUTDOWN MARGIN du ring REFUELING was categorized as an "F" (futher discussion possible) in t.he October 1-2, 1986 meeting, l f urther comparison to the GE-BWR STS Rev. 3, NUREG-0123, indicate that j acceptance of small SHUTDOWN MARGINS (LC0 3.1.1 P 3/41-1) during REFUELING with any control rod withdrawn is with the proviso that the Standby Liquid Control System be OPERABLE (LCO 3.1.5, P 3/4 1-19). The GE-BWR STS is used for comparison rather than the W-STS as the Ft. St. Vrain reactivity j control system of control rods and reserve shutdown material is similar to l l

the GE reactivity control system of control rods and Standby Liquid Control System, with neither having routine boration reactivity control as in the Westinghouse system. Also, for Ft. St. Vrain, the Actions b.2 for SHUTDOWN j and C.2.b f or REFUELING with inadequate SHUTDOWN MARGIN (LC0 3.1.3 of l November 30, 1985 Draft) require actuation of sufficient reserve shutdown material to achieve the required SHUTDOWN MARGIN. Yet in the Reserve l Shutdown System Specification (LC0 3.1.6 of the November 30, 1985 draft) l there is no requirement for OPERABILITY in either the SHUTOOWN or REFUELING condition.

LC0 3.2.1-1 i

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Please expand the Basis to describe briefly what activity is involved {

in the " determination by evaluation." What reactor observables are !

t evaluated?

LCO 3.3.2.3-3 4

The Licensee should add a calibration requirement in SR 4.3.2.3.2 for seismic instruments determined to be out of calibration following a seismic l

event. As indicated in the basis, P. 3/4 3-83 of the November 30, 1985 l i

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1 Draf t, the Licensee has already connitted to this (last sentence, fourth l paragraph). However, there is presently no requirement for such calibration in the surveillance. Although the STS guidance is to do such calibration within ten days, it is judged that 30 days provides for allowing monitoring of after shocks following the initial event.

Experience with other commercial plants is that the contractor / manufacturer provides on-site calibration capability. PSC should investigate such on-site service capability from their seismic instrument manufacturer so that the seismic instrun.ents remain on-site to be maximum 1y available for monitoring after shocks.

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1 LC0 3.3.2.3-4 The Licensee should specify monthly CHANNEL CHECKS for the seismic instruments in TS Table 4.3.2-2 (November 1985 Draf t, P. 3/4 3-82) rather than the proposed quarterly checks. PSC had proposed quarterly CHANNEL CHECKS as consistent with Turkey Point seismic instrument surveillance.

However, as Turkey Point specifications are the exception to the rule and as Turkey Point is situated in Seismic Zone 0 and FSV is situated in Seismic Zone 1, the Turkey Point specifications are not a reasonable comparable to use. Additionally, it is in the Licensee's interest to have the seismic instruments operable to facilitate restart if shutdown occurred. The relatively uncomplicated CHANNEL CHECKS would enhance instrument operability and thus the assessment of a seismic disturbance and the implications to plant restart capability.

l LC0 3.5.4 - New Item Please revise LC0 3.5.4 ACTION b (second part on page 3/4.5-30 of the TSUP draft). The words, " establish a backup system for fire suppression purposes within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ," need to be changed to the wording, " establish a backup system for fire suppression purposes prior to reaching a CALCULATED BULK CORE TEMPERATURE of 760"F but within a period of time not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ." The revision is necessary to assure an operable flow path to the liner cooling system (LCS) via the firewater system when the LCS is the only decay heat removal path during an interruption of forced cooling for l

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l purposes of maintenance and inspection. This change is consistent with the I NRC guidelines provided in the NRC letter of December 5, 1986, on PSC l 1

l commitments required to approve the proposed version to LCO 4.1.9 in the l 1 l l existing Technical Specifications. l l

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! LC0 3.6.4-2 Please provide additional information with regard to the source and approval of the acceptance criteria for PCRV concrete permeability and PCRV l Itner thinning as cited in the PSC markup of pages 3/4.6-39 and 3/4.6-40 of 1 the TSUP draft in Attachment 1 to P-87063. The NRC does not have a copy of the previous ISI procedures alluded to in the citations, and these acceptance criteria are not found in the recent ISIT submittals.

l LC0 3.7.2-4 l

The Licensee should provide additional clarification in the basis on l P. 3/4 7-14, third paragraph, second and third sentences, as they read:

"A 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ACTION time is provided to isolate the affected loop, in an effort to regain OPERABILITY of a second hydraulic fluid pump and/or at l 1 east one accumulator for the affected valve group. If OPERABILITY of a second hydraulic fluid pump and/or at least one accumulator is not restored !

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , reactor shutdown is required with 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ."

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Since OPERABILITY of a second hydraulic fluid pump is not mentioned in the Action Statement, only " supply of at least 2500 psig", the basis should use the terminology of the Action Statement, namely, supply of at least 2500 psig. Also, the basis has interpreted the Action Statement to allow one hour to restore the required conditions of at least one accumulator and or at least 2500 psig pressure before reactor shutdown within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The Action Statement a. of P. 3/4 7-12 does not address a restoration time. It requires isolation of the affected loop in one hour and reactor shutdown in the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months without regard to any restoration time.

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. o LC0 3.7.6.3-3 for the halon system in Building 10, please propose specifications consistent with SR 4.7.6.3 in the WNP No. 2 Technical Specifications, NUREG-1009 (see attached page). At WNP No. 2, tank " quantity" is determined once per six months using the heat tape and gun method approved by the American Nuclear Insurers. However, storage tank weight umst be verified periodically. PSC needs to specify an appropriate surveillance period for verifying storage tank weight and to justif.v any period exceeding 36 months.

LCO 3.7.8-4 Please correct the misspellings in PSC's markup of TSUP draft LC0 3.7.8 ACTION b. The word, " values," is misspelled twice as " valves."

LOC 3.7.10-3 Please revise FSAR Sections 1.4 and B.S.2.7 to be-consistent with the Basis definition of " safety-related" as including Class la components. PSC expressed the desire to retain the wording in the opening sentence of the Basis as being indicative of their position with regard to the scope of the safety-related snubbers. The Basis and the FSAR need to be consistent on !

this point.

LC0 3.9.1-2 The Licensee should retain the APPLICABILITY as it was in the November 1985 Draf t or should provide additional information or changes to split out Specifications 3.9.la and b from Specifications 3.9.1c and d.

Specifications 3.9.la and b are justly applicable to only "whenever both j primary and secondary PCRV closures of any PCRV penetrations are removed" l which is PSC's proposed APPLICABILITY. However, Specifications 3.9.1 c and d on requiring two startup channel neutron flux monitors and maintaining the SHUTDOWN MARGIN requirements of Specification 3.1.3, are i

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i PLANT SYSTDt5 l HALON SYSTEMS i

t!MITING CON 0! TION FOR OPERATION 3.7.6.3 The 18 Halon systees in the PGCC units in the contro1 'roce shall be OPIRA8tt with the storage tanks having at least 95% of full charge j weight and 901 of full charge pressure. I APPt!CA81LITY: Whenever equipment protected by the Halon systees is required to be OPLRABLE. ,

ACTION: ,

a. Vf th one or more of the above required Halon systems inoperable. I within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a continuous f f re watch wf th bacitse f f re I st9pression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol. ;

b. The provisions of Specifications 3.0.3 and 3.0.4 are not app 11 cable, fi '

SURVEILLANCE RE001REMEN75 i e

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4.7.6.3 Each of the above required Halon systems shall be demonstrated l OPERA 8tt: ; ;

a. At least once per 31 days by verf fying that each valve (manual, i power-operated, or automatic) in the flow path is in Its correct j position, j

b. At least once per 6 months by verf fying Halon storage task quantity l and pressure. '

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c. At least once per 18 months by: j

1. Verifying the systes, including associated ventilation systes 1 fire dampers and fire door release mechanisms, actuates. gi manually and automatically, upon recefpt of a simulated !

actuation sfgnal, and Hi

2. Performance of a flow test through accessible headers and I nozzles to assure no blockage.

d. At least once per 36 months by verifying Malon storage tank weight. l l

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WASHINGTOM NUCLEAR - UNIT 2 3/4 7-23 'a

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e appitcable throughout the REFUELING mode. Therefore, either the original APPLICABILITY in the November 1985 Draft should be retained or 3.9.la and b !

need separated from 3.9.lc and d with different applicability statements as discussed above. ]

LC0 3.9.1-3 The Licensee should revise the wording as suggested for Action b of LC0 3.9.1 (similar wording should also be used for Action C.1 of this LCO) to:

"With one of the above required neutron flux monitors inoperable, or not OPERATING, immediately suspend all operations involving CORE l

ALTERATIONS, any evolution resulting in positive reactivity changes, or movement of IRRADIATED FUEL."

The Licensee's position that the proposed words ". . . control rod movements resulting in positive reactivity changes . . ." were intended to be specific versus the NRC recommended wording ". . . any evolutions resulting in positive reactivity changes . . ." has been accepted for the other LCOs involved, LCO's 3.6.5.1-2, 3.6.5.2-1, 3.7.1.1-2, and 3.7.9-3 (other NRC actions items from October 27-30, 1986 meeting concerning reactivity changes). For those LCO's, flexibility to change between feedwater and condensate cooldown is accepted because the positive reactivity effect of any involved cooldown has already been accounted for in the SHUTOOWN MARGIN requirement of 0.01 delta k in accordance with LC0 3.1.3. And the reactivity additions are relatively small and added slowly as PSC explained in Attachment 4 to their February 20, 1987 letter. l For the subject LCO, however, Actions b and and C.1 involve loss of one and I both startup channel neutron flux monitors, respectively. Allowing an intended positive reactivity change due to cooldown with degraded startup channel neutron flux monitoring is unacceptable. When both startup channels are inoperable, any controllable positive reactivity change should be stopped, as under these conditions, immediate assessment of flux changes is lost. With only one operable startup channel of neutron flux monitoring, it is also unacceptable to intentionally make positive 9

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reactivity changes as that one startup channel may be operating erroneously and there is no second operating channel to confirm it's readings. . Also, with only one operating channel, a sudden loss of that channel again ;

results in complete loss of the ability to make immediate assessments of neutron flux changes. These changes would make the FSV Actions consistent with those of STS Rev. 5, P. 3/4 9-2, on neutron flux monitoring capability j during refueling, and with those of the existing FSV Technical Specifications, LC0 4.7.1. J AC 6.3/6.4-1 The Licensee should include reference to the NRC March 28, 1980 letter in Technical Specification Sections 6.3.2 and 6.4.1 per the subject NRC c onnen t . Contrary to the Licensee's position in the October 27-30, 1986 ]

Meeting that the NRC March 28, 1980 letter is not in their letter log and therefore is doubtful that is was agreed to, the Licensee responded to it in their letter of December 20, 1980 (P-80438). In Attachment 1 of that l letter, the Licensee stated that although they had received the NRC I March 28, 1980 letter too late to implement the requirements by August 1, 1980 as required by the letter's Enclosure 1, some requirements would be met by the August 1, 1980 date and their program for compliance would be submitted by .lanuary 15, 1981. As it appears that the Licensee l

has connitted to the subject NRC March 28,1980 requirements, stating this '

in the Technical Specifications should not involve any undue hardships.

AC 6.5.1.6a (AC 6.5-1)

The Licensee should revise AC 6.5.1.6a to require the PORC to review any procedures required by AC 6.8.4 as AC 6.8.4 includes procedures for lodine sampling in the reactor building and Post-Accident Sampilng, both of 1 which are TMI-2 Action items of NUREG-0737 and Generic Letters 83-36, 37.

Also, the Licensee should revise AC 6.5.2.7 to require the NFSC to review the programs of AC 6.8.4. This is what PSC said PORC and NFSC do now (See PSC letter of February 20, 1987, Attachment 2). However, as written, AC 6.5.1.6a and AC 6.5.2.7 do not have these requirements but should, 10

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j AC 6.5.1.7b (AC 6.5-1)

The Licensee should provide additional information to explain what

" Pro":!ure Deviation Reports" are, as used in their letter of February 20, 1987, Attachment 3, Item AC 6.5# 1-7 and Attachment 1, )'

P. 6-15, marked up item 6.5.1.7.b. In Attachment 3, reference is made to

' Temporary Changes" whereas in Attachment 2, reference is made to

" Procedure Deviation Reports". It is not clear if these are one and the same thing. If they are the same, PSC should provide additional information to justify the proposed exception for " Temporary Changes".

AC 6.9.1.2.a (AC 6.9-1)

The Licensee should revise their proposed marked up P. 6-27 (PSC j 1etter of February 20, 1987, Attachment 1) to either put report submittal ]

times in 6.9.1.2, first paragraph as in the STS Rev. 5, P. 6-16, or place a l report submittal directly in 6.9.1.2.a.2. As proposed, 6.9.1.2.a.2 does not have any report submittal time connected to it. PSC placed their report submittal directly in 6.9.1.2.a.1 rather than 6.9.1.2. Therefore, when 6.9.1.2.a.2 was added, it was not covered by a report submittal time.

The following RAI is a result of the review of Attachment 1 to P.-85448 NRC Action 3 PSC should provide the Environmental Qualification study for this item.

The following is NRC Action Item 3 and its response from Attachment 1 to PSC letter of November 27,1985 (P.- 85448) .

NRC Action Provide guidance on whether components, which are required to function to maintain other equipment within an environment for which it is qualified, should also be in the Technical Specifications (for example, main steam isolation valves).

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1 NRC Response Where assumptions for equipment operability related to environmental qualification are based on the successful operation of active components in the event of an accident, the availability and reliability of these l components should be ensured through Technical Specification requirements.

l Specific items of concern identified by the staff were the main steam isolation valve (MSIV) operability and closing time requirements as well as the hot reheat (HRH) valve operability requirements. Other EQ operability I requirements should be identified by the licensee and incorporated as Technical Specification requirements to ensure equipment necessary to mitigate accidents function within the assumed environmental conditions.

In addition to the EQ analysis, other analysis which rely on equipment i

operability may also result in Technical Specification requirements, for example, the equipment operability requirements based on previous fire protection analysis (Appendix R Evaluation: FSV Reports 1 through 4) should be reflected in the Technical Specifications.

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. - ENCLOSURE 3 Additional Comments on The Fort St. Vrain Technical Specification Upgrade Program Final Draft LCOs for Safety-Related Cooling Functions l

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NRC COMMENTS-LCO 3.5.1.1

1. Based on the PSC letter (P 87002) dated January 15, 1987, the condition statement for LCO 3.5.1.1.a.2 needs to be rewritten as follows:

Both steam generator sections (both the economizer-evaporator. 3 superheater (EES) and the reheater) OPERABLE including two OPERABLE flow I paths. l

2. Previously, the safe shutdown cooling outlet flow paths were via the by-pass valves off each loop's superheater outlet and each loop's hot reheat steam line. The by-pass valves were verified to be OPERABLE as part j

, of the normal operation of the bypass function. The recently installed six l l

inch vent lines described in P-87002 are apparently not to be used on a i routine basis. Therefore, SR 4.5.1.1.b needs to be revised by renumbering l surveillance b.2 and b.3 to b.3 and b.4, respectively, and adding a new surveillance as SR 4.5.1.1.b.2. The new SR 4.5.1.1.b.2 should read as follows:

At least once per 18 months by verifying the OPERABILITY of each I superheater outlet flow path by verifying that the valves in the six inch vent lines can be opened and that the vent flow paths are not obstructed.

3. Subject to NRC final approval of the proposed revisions to the Basis for the existing LCO 4.1.9, the following paragraph needs to be added at the bottom of the fourth page of the Basis for LCO 3.5.1.1 following the second paragraph of the subsection entitled Redundancy Criteria:

Specification 3.0.N provides the methodology and necessary data to determine the appropriate time interval to reach a CALCULATED BULK CORE TEMPERATURE of 760 degrees F. If the active core remains below this temperature, which corresponds to the design maximum core inlet temperature as indicated above, then the design core. inlet temperature cannot be exceeded and there can be no damage to fuel or PCRV internal components regardless of the amount, including total absence, or reversal, of primary coolant helium flow.

PSC needs to provide the appropriate number "N" for the cross-referenced specification. The need for Specification 3.0.N has been identified in the NRC Request for Additional Information at Enclosure 1.

4. The first paragraph of the subsection entitled Steam Generators on the fifth page of the Basis for LCO 3.5.1.1 and LCO 3.5.1.2 needs to be replaced with the following paragraph.

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, 2 Whenever the CALCULATED BULK CORE TEMPERATURE exceeds 760 degrees F, I both the reheater and EES sections of the steam generator must be OPERABLE. The steam generator reheater or EES sections can receive j l water from -cither the emergency condensate header or the emergency feedwater header as required to be OPERABLE per this Specification and per Specification 3/4.5.3. System' flow OPERABILITY is determined by verifying flow from each of the aforementioned emergency headers (see LCO 3.5.3.1) through each section of each steam generator. Whenever the CALCULATED BULK CORE TEMPERATURE is less than or equal to 760 degrees F or the plant OPERATIONAL MODE is REFUELING, system flow OPERABILITY is determined by verifying flow from either of the aforementioned emergency headers (see LCO 3.5.3.2) through either section of either steam generator, j

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5. An additional paragraph is also required under the subsection entitled Steam Generators on the fifth page of the Basis for LCO 3.5.1.1. This paragraph needs to discuss the appropriate operability requirements for the {

seismically and environmentally qualified six inch vent lines and to cite the supporting safety analysis requiring the use of these vent lines.

PROPOSED RESOLUTIONS 1./2./3./4./5. PSC needs to incorporate the required changes.

NRC COMMENTS-LCO 3.5.1.2

1. Contrary to NRC Comment No. 1 on LCO 3.5.1.2 as given in Enclosure 1 to the NRC letter, Heitner to Villiams, April 17, 1987, the words to be added after the word "0PERABLE" in the condition statement for LCO 3.5.1.2.a.2 need to be " including one OPERABLE flow path," not two. LCO 3.5.3.2 requires only one flow path, either the emergency condensate header or the emergency feedwater header, to be OPERABLE in STARTUP and SHUTDOWN l

whenever the CALCULATED BULK CORE TEMPERATURE is less than or equal to 760 degrees F or in REFUELING.

2. See NRC Comment No. 4 on LCO 3.5.1.1.

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3. Delete LCO 3.5.1.2.b.1 and renumber b.2 through b.5 as b.1 through b.4, respectively. The deleted LCO is neither needed nor appropriate since the boiler feed pumps are not required to be OPERABLE under the same APPLICABILITY statement (see LCO 3.7.1.1) nor necessarily the emergency i feedwater header (see LCO 3.5.3.2). For decay heat levels below that for I shutdown from 35% power, previous analysis has shown that fuel damage will not occur in the depressurized core as long as one train of the PCRV LCS is operating. Necessary modifications to the TSUP draft to incorporate the guidelines from the NRC letter, Heitner to Williams, dated December 5, 1986, are indicated elsewhere in these comments, and these modifications are directed to providing assurance that the PCRV LCS provides a backup cooling capability during the conditions of APPLICABILITY. However, NRC reserves the right to modify LCOs 3.5.1.2, 3.5.3.2, and 3.7.1.1 pending completion' of the review on the occurrence frequency of rapid depressurization events.

J'

, 3 In addition, further modifications to Specifications 3/4.5.1, 3/4.5.3, and l - 3/4.7.1.1 may be required depending upon the final resolution to NRC-Comment No. 1 on Section 3/4.4 as documented in Enclosure 1 to the NRC letter, Heitner to Williams, April 17, 1987. As noted in the subject I

comment, the Updated FSAR does not support operation with the reactor producing fission heat simultaneously with reactor cooling using condensate only. When fission heat is being produced'in the critical core, normal cooling should be provided by forced circulation using either steam drive j or feedwater-drive of the helium circulators. The current wording of Specifications 3/4.5.3 and 3/4.7.1.1 do not provide for the OPERABILITY of the feedwater-drive for the safety-related emergency core cooling function when the. CALCULATED BULK CORE TEMPERATURE is less than 760 degrees F although fission heat is allowed to be as high as 5% of rated reactor power in STARTUP. If new specifications are not added to address the OPERABILITY of the feedwater drive for the important-to ssfety normal cooling function whenever fission heat is being produced, further restrictions may be required. 1 One option is to delete the footnote on STARTUP in LCOs 3.5.1.1, 3.5.3.1, and 3.7.1.1 and to delete STARTUP in the APPLICABILITY in LCOs 3.5.1.2 and 3.5.3.2. Alternately, to allow the flexibility of performing training s arts 'while limiting the allowed fission heat level, the footnotes on the i subject LCOs can be changed as follows. In LCOs 3.5.1.1, 3.5.3.1, and 3.7.1.1, the footnote should be changed to read: "Whenever CALCULATED BULK CORE TEMPERATURE is greater than 760 degrees F, or, in STARTUP, whenever reactor thermal power is equal to or greater than 2 percent of rated reactor power." In LCOs 3.5.1.2 and 3.5.3.2, the footnotes should be changed to read: "Whenever CALCULATED BULK CORE TEMPERATURE is less than or equal to 760 degrees F, or, in STARTUP, whenever reactor thermal power is less than 2 percent of rated reactor power." In addition, the Bases for the affected LCOs should be modified to indicate that the limits on STARTUP are to allow flexibility for achieving criticality during training starts and that procedures have been implemented to limit the amount of fission heat to much less than 2% during such activity. ,

4. LCO 3.5.1.2 ACTION b needs to be rewritten as follows;

b. With less than the above required OPERABLE equipment and with no forced circulation being maintained, be in at least SHUTDOWN within 10 minutes and suspend all operations involving CORE ALTERATIONS, control rod movements resulting in positive reactivity changes, or movement of IRRADIATED FUEL, and either:

1. Restore forced circulation on at least one loop prior to !

reaching a CALCULATED BULK CORE TEMPERATURE of 760 degrees F, :

and comply with ACTION a, or, J

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

L.

i

. 4 '

5. In LCO 3.5.1.2, both ACTIONS a and b, the words " CALCULATED BULK CORE TEMPERATURE" are used with regard to the required restoration of equipment or conditions. In both instances, a footnote symbol should be added after !

the word " TEMPERATURE" with the following words provided in the text of the footnote:

Specification 3.0.N provides the methodology and necessary dats. to determine the appropriate time interval to reach a CALCULATED BULK CORE TEMPERATURE of 760 degrees F. 1 I

PSC needs to provide the appropriate number "N" for the cross-referenced specification. 1 PROPOSED RESOLUTIONS j I

1./2./3./4./5. PSC needs to incorporate the required changes, l 1

3. PSC needs to provide additional specifications per the previous cited comment on Section 3/4.4 or to incorporate the suggested changes in the LCOs cited or to propose and justify alternatives. ,

l

-1 NRC COMMENTS-LCO 3.5.3.1 1

1. In SR 4.5.3.1, the reference to Specifications 4.5.2.1.a needs to be deleted and replaced with a reference to Specification 4.5.1.1.b.1.

2. See above NRC Comment No. 3 on LCO 3.5.1.2.

PROPOSED RESOLUTIONS

1. PSC needs to incorporate the required changes.

2. Same as above for NRC Comment No. 3 on LCO 3.5.1.2.

NRC COMMENTS-LCO 3.5.3.2

1. The ACTION for LCO 3.5.3.2 needs to be deleted and replaced as follows:

With both the emergency feedwater and emergency condensate header inoperable, be in at least SHUTDOWN within 10 minutes and restore at least one header to OPERABLE status prior to reaching a CALCULATED BULK CORE TEMPERATURE of 760 degrees F and suspend all operations involving CORE ALTERATIONS, control rod movements resulting in positive reactivity changes, or movement of IRRADIATED FUEL.

2. In the ACTION, PSC should add the same footnote as described above in NRC Comment No. 5 on LCO 3.5.1.2. Similar information and cross references as provided in the footnote should also be included in the Basis.

3. In SR 4.5.3.2, the reference to Specification 4.5.2.1.a needs to be deleted and replaced with a reference to Specification 4.5.1.1.b.1.

4 See above NRC Comment No. 3 on LCO 3.5.1.2.

l 1

l

4 5

1 PROPOSED RESOLUTIONS 1

1./2./3. PSC needs to incorporate the required changes.

4. Same as above for NRC Comment No. 3 on LCO 3.5.1.2.

NRC COMMENT-LCO 3.5.4 In ACTION b (the second part of the ACTION statement on page 3/4 5-30), PSC should add the same footnote as described above in NRC Comment No. 5 on LCO 3.5.1.2. A request to modify this ACTION has been provided in the NRC Request for Additional Information at Enclosure 2. The words "within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months " need to be replaced with the words " prior to reaching a CALCULATED ,

BULK CORE TEMPERATURE of 760 degrees F but within a period of time not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ." I Similar information and cross references as provided in the footnote should also be included in the Basis. l PROPOSED RESOLUTION PSC needs to incorporate the required changes.

NRC COMMENTS-LCO 3.6.2.2

1. In LCO 3.6.2,2, the condition statement should be deleted and replaced with the following:

The Reactor Plant Cooling Water (RPCW)/PCRV Liner Cooling System (LCS) i shall be OPERABLE with: I

a. One RPCW/PCRV LCS loop OPERATING with at least one heat exchanger and one pump in each loop OPERATING, and

b. With firewater supply available via one OPERABLE flow path. !

The change is necessitated to comply with the NRC guidelines for PSC commitments with regard to proposed revisions to existing LCO 4.1.9. These !

guidelines are given in the NRC letter, Heitner to Williams, dated i December 5, 1986. '

2. In LCO 3.6.2.2, the ACTION statement should be deleted and replaced with the following:

With no RPCW/PCRV LCS loop OPERATING. within 10 minutes, be'in at least SHUTDOWN and suspend all operations involving CORE ALTERATIONS, control ;

rod movements resulting in positive reactivity changes, or movement of '

IRRADIATED FUEL, and:

a. Restore at least one loop to OPERATING status prior to reaching a CALCULATED BULK CORE TEMPERATURE of 760 degrees F, or

b. Restore forced circulation cooling prior to reaching a CALCULATED' BULK CORE TEMPERATURE of 760 degrees F.

\;

i 6

The change is necessitated to comply with the NRC guidelines for PSC commitments with regard to proposed revisions to existing LCO 4.1.9. These guidelines are 61 ven in the NRC letter, Heitner to Williams, dated December 5, 1986.

3. If the interfacing isolation valves between the firewater system and the RFCW/PCRV LCS are not covered in the survefilances on the SAFE SHUTDOWN COOLING water supply system per SR 4.5.4.1.f or SR 4.5.4.1.g.3, the subject isolation valtes need to be covered by revising SR 4.6.2.2 appropriately.

4 In the ACTION, PSC should add the same footnote as described above in NRC Comment No. 5 on LCO 3.5.1.2. Similar information and cross references as provided in the footnote should also be included in the Basis.

l PROPOSED RESOLUTIONS 1./2./3./4 PSC needs to incorporate the required changes. i NRC COMMENT-LCO 3.7.1.1 See above NRC Comment No. 3 on LCO 3.5.1.2.

PROPOSED RESOLUTION Same as above for NRC Comment No. 3 on LCO 3.5.1.2.

NRC COMMENT-LCO 3.7.1.6 In the ACTION, PSC should add the same footnote as described above in NRC Comment No. 5 on LCO 3.5.1.2. Similar information and cross references as provided in the footnote should also be included in the Basis.

l PROPOSED RESOLUTION PSC needs to incorporate the required changes. )

I NRC COMMENT-LCO 3.7.4.2 In the ACTION, PSC should add the same footnote as described above in NRC Comment No. 5 on LCO 3.5.1.2. Similar information and cross references as l provided in the footnote should also be included in the Basis. l PROPOSED RESOLUTIONS PSC needs to incorporate the required changes.

l

%- Amendment No. ENCLOSURE l Page 3/4.5- .

SAFE SHUTDOVN COOLING SYSTEMS bi$/YT -

2 3/4.5.1 SAFE SHUTDOWN COOLING EQUIPMENT ,

1 LIMITING CO OITION FOR OPERATION l

1

\

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

1

1. One Ifue circulator OPERA 8LE, and 1

2. steam generator.ection[( he economizer-evaporizer-superheater (EES) g the reheater)

OPERABLh pcl I l ' " '

D ** b. For OPERABLE helium circulators, the following safe.

OPERAELE shutdown cooling drives and auxiliary equipment shall be OPERABLE:

St., p ot k

1) A safe shutdown' cooling drive with the capability of providing the equivalent of 8000 rps circulator speed at atmospheric pressure to two circulators e simultaneously, rO[\ A p Y#

gpC V n 2) Jw6 safe shutdown cooling drivef with the capability of providing 3% ra.ted helium flow at operating g *, M Q pressure with firewater supply, including two d\

Os -

OPERABLE emergency water booster pumps and oeERABLE fiow a ths.

two

' 3) The turbine water removal system shall be OPERA 8LE, 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

l

. a te.endrent No.'

Page 3/4.5-DQ;* .T - l v

FEB o . -

ng

[] b) At least 10% of primary coolant pressure w e boundary bolting and other structural bolting W which has been removed for the inspection above g and which is exposed to the primary coolant M

shall be nondestructively tested r g for identification'of inherent or developed defects. 1

- " y c) Reports lL w

O f Within 90 days of examination completion, a N Sh special Report shall be submitted to the NRC in accordance wtth spect fIcation 6.9.2.

4. This

& I report shall include the results of the helius circulator examinations.

S ne. b. The steam generators shall be d:monstrated OPERA 8LE:

l 4 y 1. At .least once per 18 months be n,

8 through the emergency feedvater/header verifying proper flow and emergency

) , ,

y condensate header to the steam generator sections.

}

l J

f t

g/. At*1 east once per 5 years by volumetrically examining

v 'r D the accessible portions of the following bimetallic wg for indications of subsurface defsets:

welds "w3f .

Y) The . main steam ring header collector to

,e ! , collector drain piping weld for one steam generator module in each loop, and l

f ..f n{ c. ,8) ' The same two steam generator modules shall be re examined at each interval, v a. w

.c "

F 0 .f The initial examination shall be performed during SHUTDOWN or REFUELING prior to the beg. inning of

, cycle 5. fuel This initial exastnation shall also include the btmetallic welds described above rcr two additional steam generator sodules in each loop.

J/, /. LubeLeakExamination .

l Each time a steam generator tube plugged due to a 1eak, specimens from the accessible subheader tubes connected 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 a11 preytous tube itaks.

At ta a,1 o n a a p a r- te en..rhs by ver sry i,, tkD O PE R A B I.L.I.T Y o f ea t h suret h eale r- estia (t.w p ath b yv e r- ; 4 y 'e n ) .T h o r c he valves I ., t h e. six ihe b ve_nT lines can b e o p en e d An s( Th s T T4 C et .f. [ o w p a'T h r Jre h e't o b r T r o cT e d l

1

' no Amendment No.

Page 3/4.5-

. l

$AFE SHUT 00VN COOLING SYSTEMS 3/4.5.1 SAFE SHUTDOWN COOLING EOUIPMENT p.q f, h-FEB2 m-LIMIT!NG CONDIT10N FOR OPERATION.

)

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

1. One helium circulator OPERABLE, and

2. One steam generator section (eMer the econostzer-

, g , j ,.

evaporator-superheater (EES) or reneater) OPERABLp one O URABLE b. For at least one OPERABLE helium circulator, the following i

, flow f *h emergency drives and auxiliary equipment shall be OPERABLE: ;

I .g sa fe shutdown cooling drive with the capability of providing with 3% rated heltum flow at opera ting pressure firewater supply, including one OPERABLE emergency water booster pump and one OPERABLE flow path, j

Ig. The turbine water removat system, including one turbine water removal pump, FOR

2 /. The norr.ai bearing water .xstem, inciuding one source or

'{t,[ y,g bearing water makeup and one bearing water makeup pump (P2105 g P2108), and j

{

'l g.

O N* . *

~ The bearing water accumulator (T-2112 Tt2113. T-2114, g T-2115) for the OPERABLE circulator (s).

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

., .e Amendment No.

page 3/4.5-bbe17' l ACTION: a. With less than the FEB 3 A *- !

with forced circulation maintained, be in atabove required OPERA wgy, eo L. m '. wii, and least SHUT 00VN

. restore the required equipment to OPERABLE

" ' " I ' ' status prior to reaching a CALCULATED BULK CORE TEMPERATURE K of 760 degrees F. or suspend all operations involving CCRE ALTERATIONS,. . Cont,.o/ rodneovernenTs resulting in positive re' activity changes, or, movement of 3RRADIATED FUEL.

<Ng. [ Dv eq 1 t L q pm t 4 n ar op n t 4

.e4 t nd e t s o p t E gre s , r

/>s/ // u t i es u ,v o As es[orMvhebfI IA ,

ta

/

2.

'\ '

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

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

1 !

b.

With less than the above required OPERABLE forced equipment and with no circulation being maintained, be in at least SHUTDOWN within fo minutes control rodand suspend all operations involving CORE ALTERATIONS, i movement of IRRADIATED FUEL, and either: movements resulting in positive re 1.

Restore forced circulation on at least one loop prior to reaching a and comply with ACTION a, orCALCULATED BUlX CORE TEMPERATUR gc I, pre. .d C w p "-

/

Specification 3.0. ovides the methodology and necessary data to determine the appropriate time interval to reach a CAIEULATED BUlX CORE TEMPERATURE of 760 degrees F.

Amendment No.

Page 3/4.5-Depressurization '

min g g g . ,;.g in the unitkely event that all forced cireplation is lost for 90 minutes, start of depressurization is initiated as a function of prior power levels, with 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months from 1005 RATED THERMAL POWER being the most limiting case. Opera tors will continue attempts to restore forced circulation cooling until 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after the loss of forced circulation. Multiple sources and flowpaths to establish forced convection cooling using circulators makes required depressurization highly unifkely. Cooldown using forced circulation cooldown is preferred to a depressurized cooldown, with the PCRV liner cooling system. Depressurization of the PCRV under extended loss of forced circulation conditions is accomplished by venting the reactor helium through a train of the helium purification system and the reactor building vent stack filters to atmosphere. Start of depressurization times from various reactor power conditions are delineated in Figures 3.5.1-1, 3.5.1 2, and 3.5.13 and are discussed in the FSAR Section 9.4.3.3 and Appendix 0.

Redundancy Criteria l

The use of 760 degrees F CALCULATED BULK CORE TEMPERATURE as a division between the APPLICABILITY of Specification 3.5.1.1 1

verses 3:5.1.2 is explained as follows:

In the FSV HTGR, the limiting parameter of interest is a core inlet temperature greeter than 760 degrees F. The CALCULATED BULK CORE TEMPERATURE is a conservative calculation of the maximum potential temperature in the core and surrounding components. The conservatism 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 l degrees F, it is acceptable to require only one OPERABLE system for accident prevention and mitigation without single failure consideration, on the basis of the Itmited core cooling requirements. ;

nad P S C. ~C o p r o A hYI W

Specification 3.0hprovides the methodology and necessary data to determine the appropriate time interval to reach a CA14U1ATED BULK CORE TEMPERATURE of 760 degrees F. If the active core remains below this temperature, which corresponds to the design maximum core inlet !

temperature as indicated above, then the design core inlet temperature cannot be exceeded and there can be no damage to fuel or PCRV internal components regardless of the amount, including total absence, er reversal, of primary coolant helium flow.

e Amendment No.

. Page 3/4.5- l DRAFT 4

i All forced circulation may be interrupted for maintenance hb, CORE TEMPERATUREpurposes provided that the time calculated for CALCUL! 1 to reach 760 degrees F is not exceeded.

However, if forced circulation is temporarily res tored, a recalculation shall be performed, based on present conditions, to estabitsh 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 that forced circulation time to reach CALCULATED SULK CORE TEMPERATURE is maintained. Theequal degrees F may be recalculated as of ten as required. to 760 Steam Generators DeseT <

2d Th st d g er rehe om the the s cia ed EES gen cti ns an ec ve t f5 C n u //

Reptac^ em e ater ead wh en es r s

per are r air T . p r ..., J c. {

w.th. i pec at n. e b

stem P ITY s te g g ,, _

ve ify (1 fr ch of t ef ead st o h ch am g ra r for e on d r yy ytragraph i t

en .c i t i < h h Bimetallic Weld Examination ' '

The steam generator 2nd crossover tube bimetallic welds between pg eT.' n3 )

Incoloy 800 examination. and 21/4 Cr 1 Mo materials are not accessible for The bimetallic welds between steam generator g g'7 'l ring header collector, the main steam piping, and the 2waty 3 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 geunetrically 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 at thematerials. ' .

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

) _

Vhenever the CALCULATED BULK CORE TEMPERATURE both the reheater and EES sections of theexceeds OPERABLE. steam 760 degrees F, generator must be The steam generator reheater water from either the emergency condensate or EES sections can receive I feedwater header as required to be OPERABLE per thisheader or the emergency per Specification 3/4.5.3. System Specification and ifying flow from each of the aforementionedflow OPERABILITY is determined by Lc0 3.5.3.1) through each section of each steam generator. emergency headers (see Whenever the CALCULATED or the BULE CORE TEMPERATURE is less than or equal s F to 760 degree plant OPERATIONAL MODE is REFUELING, system flow OPERABILITY is i

determined headers (see LCO by verifying 3.5.3.2) flow from either of the aforementioned ncy emerge generator. through either section of either steam k -

J .

,.e.,:.

- Page 3/4 5-26

DRAFT SAFE SHUT 00WN COOLING SYSTEMS 3/4.5.3 EMERGENCY CCNDENSATE AND EMERGENCY FEE 0 WATER WEACE8 EMERGENCY CONCENSATE AND EVERGENCY FEECWATER 9.,g WEACERS ~ w.. - ur:

LIMITING CCNOIT*0N F 9 CoERATICN 3.5.3.1 The emergency :endensate header and the emergency feedwater header shall te CPERABLE.

APPLICABILITY: POWER, LCW PCWER, STARTUP*, and SHUT 00WN" ACTION:

Vith either the emergency concensatt header or the emergency feedwater P.eader inoperable, restore the inoperable header to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or:

1. When in POWER, LOW POWER, or STARTUP, be in at least SHUTOOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or

2. When in SHUT 00WN, suspend all operations involving control rod movement resulting in positive r.e activ i ty changes, or movement of IRRADIATED FUEL.

'-~,

)

SURVEILLANCE REOUIREWENTS 4.5.3.1 No accitional Surveillance Requirements are required other tnan those surveillance identified in Specification " '. .I 1

9, s,1, I, b . I 1

Whenever degrees F. CALCULATED BULK CORE TEMPERATURE is greater than 760 1

l

~

t . 8-- ,3,

, Page 3/4 3 27 .

DRAFT SAFE SHUTOCWN COOLING SYSTEMS lQy J Q gg j i

3/4.5.3 EMERGENCY CCNOENSATE AND EMERGENCY FEE 0 WATER HEADERS EMERGENCY CONDENSATE AND EMERGENCY FEE 0 WATER HEACERS .. - SE ..IM l

LIMITING CCNCITION FOR CPERAT!GN I

3.5.3.2 Either the condensate header or the emergency feedwater header shall be OPERABLE.

APPLICABILITY: STARTUP", SHUT 00VN", and REFUELING -

ACTION: With o t, th er ncy f dwat and gency h der opera e, densa \

estor a e st one a r 0 LE sta s to to *1m cale j fremx ecay he ed fo the c to he t ' 1 to each g CULATED ' LK COR TEMP nA U' 760 deg j in l y i '- CC c A F an spen all e ations i i

3 TIONS o ent I ro vement es ting j pq iti reac ty chang s, or move ent or RRAOIATED EL/ ,

l SURVEILLANCE RE0VIRE9ENTS x- {

4.5.3.2 No additional J Surveillance Requirements are required other than those surveillance identifiec in Specification

-A 5 L 1 . .O q , y ,1.1. b . I I

l

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

/ PSC- mced5 TPm'J*T~

Specification 3.0hprevides the methodolo6y and necessary data to determine the appropriate time interval to reach a CALCUIATED BUlX CORE TEMPERATURE of 760 degrees F.

t g_ _ -

} With both the emergency feedwater and emergency condensate header inoperable, be in at least SHUTDOWN within 10 minutes and restore at least one header to OPERABLE status prior to reaching a CALCUIATED BULK CORE TEMPERATURE # of 760 degrees F and suspend all operations involving CORE ALTERATIONS, control rod movements resulting in positive reactiviry changes, or movement of IRRADIATED FUEL.

~

\

cer.:..e t so.

Page 3/4 5-28 -

DRAFT BA515 FOR SPECIFICATION LCO 3.5.3 / SR 4.5.3 NOV 3 0, W The OPERABILITY of the emergency condensate header and the emergency feedwater header ensures redurcant water supoly paths to the helium circulators and steam generators for SAFE SHUTOOWN COOLING of the plant.

In th e* e v en t"Tf~"a

failure of the normal. feedwater lide, the availability of either the emergency feedwater or emergency condensate lines provides aceouate shutcown capacility. CPERABILIIY of the aforementioned headers is accomplished d; ring SHUT 00WN by verifying flow through each header to the steam generators and helium circul,ators.

[In the FSV HTGR, the limiting parameter of interest is~ a f core inlet temperature greater than 760 degrees F. The CALCULATED BULK CORE TEMPERATURE is a conservative calculation of the maximum potential temperature in the core and surrounding components.

The conservatism are such that if the degrees CALCULATED BULK CORE TEMPERATURE is limited to 760 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 P S C- criterionthen greater whenever CALCULATED 760 degrees F. However, BULK CORE TEMPERATURE is when CALCULATED BULK h e.ed s CORE TEMPERATURE is equal to or less than 760 degrees F, it is acceptable to require only one OPERABLE system for Io accident prevention and mitigation wit sut single failure N

consideration, on the basis of tne limited core cooling cross ; requirements, e e4erene All fo'rced circulation may be inter .:ted for maintenance SP uhab

purposes provided that the time calculated for CALCULATED BULK CORE TEMPERATURE to reach 760 degrees F is not gg exceeded. However, if forced circulation is temporarily restored, a recalculation ; hall be performed, based on present conditions, to establish a new time period for CALCULATED BULK CORE TEMPERATURE to each 760 degrees F.

Redundant systems may also be taken out of service for maintenance or surveillance testing previded that forced circulation is maintained; and the time to reach CALCULATED-BULK i

CORE TEMPERATURE equal

( recalculated as often as required. to 760 degrees F may be The emergency feedwater header is not normally placed in service until approximately 30% reactor ocwer, to prevent unnecessary long-term wear of components associated with the emergency feedwater header.

Nevertheless it is still required to be OPERABLE during the aforementioned 1400ES.

_ _ . . _ . _ _ _ _ _ m_ _ ._ _.___.m_m._u__.___.m___m.__._m__m.___s

~... . .... e - t N :

. - Page 3/

It0V 3 0 885 i With CALCULATED degrees F: BULK CORE TEMPERATURE less than or equal'to 760 I

a. With one of the above recuired pumps and/or makeup conds -

fnoceracle, restore the inoperable e:ulpfe*nt to status with CFERABt.E or water succly.14 days or provide an alternate backup puc.o The provisions of Specification 3.0.4 are not applica3'e.

b. I With the SAFEestablish inoperacle, SHUTCOWN COOLING water system otherwise '

a backup system for suppression purposes M

a M t a s .2- fire ,

t !

1 SURVEILLANT REQU!REMENTS

\

4.5.4.1 The SAFE SHUT 00WN COOLING water supply system shall be demonstrated OPERABLE:

a.

At least once per 7 days by verifying the contained water

- supply volume in each of the circulating water makeup ponds, l

b.

At least once per 31 days by starting the electric motor-driven fire pump and operating it for at least 15 i minutes. l a

, f c. At least cnce water makeup pump that is not alrea:/ running.per 31 days d.

j At'least once per 31 days by verify'r.g that each valve in the flow path, that is not locked, sealed, or otherwtse secured in place is in its correct position,

e. At least once per 12 months by performance of a system flush, f.

At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel, g.

At least once per 18 months by performing a system functional actuation test which includes simulated automatic of the system throughout its operating secuence, and:

1. Ve ri fying that the automatic valve in the flow path actuates to its correct position. ,,

2. Verifying that each pump (motor-drAi/en and engine-driven) develops at least 1425 som at 119 psig, prior to within reaching a period a CALCULATED of time not to exceedBULK CORE TEMPERATUREkf 760 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . but degrees F P 5 c % ,

d s T . P r. g i s c M ",

g Specification 3.0 provides the methodology and necessary data to determine the appropriate time interval to reach a CALCULATED BU1K CORE TEMPERATURE of 760 degrees F.

O

3 I~D P y NOV 3 0 two With two out of three circulating water makeup pumps inoperable or with any one firewater pump circulating water makeup pond inoperable or a inoperable and the C4LCULATED BULK CORE ' i TEMPERATURE greater than 760 degrees F, a restoration .

time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and pumps. is consicered sufficient based on redundant flow patns However, with all circulating water makeup pumps, hencers, or firewater pumps inoperable, a restoration time of I l hour supplyis specified, are lost. as all means of SAFE SHUT 00WN COOLING water i 1

The surveillance that all eouipment, water supplies, and flow paths remain willidentified OPERABLE as specified in order to meet those SAFE SHUT 00wH COOLING requirements specified above. ~

The use of 760 degrees F CALCULATED BULK CORE TEMPERATURE as a division between the ACTIONS is explained as follows:

In the F5V HTGR, the limiting parameter of interest is a core inlet temperature greater than 760 degrees F. The CALCULATED BULK CORE TEMPERATURE is a maximum potential temperature conservative calculation of the P5C components. in the core and surrounding i h##g# , The censarvatisms are suen that if the CALCuuTED BULK CORE TEMPERATURE is limited to 760 degrees F, E

inlet temperature of 760 degrees F is not exceeded. the design 7' Systems used satisfy forthe accident prevention and mitigation are required to

( c_ r o i s -

CORE TEMPERATURE is greater snanHowever, 760 cegrees' F. singl r e fe re n c e

/ CALCULATED BULK when CORE TEMPERATURE is ecual to or less than 760 cegrees for accident F, it is acceptable to require oniy one OPERABLE system 5f # cJJte **"5'**" " prevention" *"' D5 and mitigation without single failure requirements.' ' th' ii*d d

*I'"9 2'*

All forced circulation may be interrupted for maintenance

] purposes TEMPERATURE provided that the time calculated for CALCULATED BULK to reach 760 degrees F, is not exceeded. However, j] If forced circulation is temocrarily restored, a recalculation

snall be performed, based on present concisions, to establish a new time period for CALCULATED BULK CORE TEMPERATURE 4 760 degrees F. to reach t Redundant systems may also be taken out of service for maintenance or surveillance testing provided that forced circulation is maintained; and the time to reach

,(CALCULATEDBULKCORETEMPERATUREequalto760degreesFm recalculated as often as required. be

- - _ ~ _ - - - - _ _ - - _ _ - - - _ . - - _ _ _ _ - _ - - - - _ _ _ _ - - . - - - _ _ _ _ _ - _ . _ _ - _ . . _ - . - - .- _ - . . - __n- - - - - - _

~

Amendment No.

Page 3/4 6-DRM:-

PCRV AND CONFINEMENT SYSTEMS _

F E B 2 e '"cc a 3/4.6.2 REACTOR PLANT COOLING VATER /PCRV LINER COOLING SY

$HUTDOWN LIMITING CONDITIONS FOR OPERATIONS 1 i

3.6.2.2 a n P) R er o b 8*'

Il e a 1 s on RP R C

A i

e e ian r n n p in TI .

APPLICABILITY: STARTUS SHUTDOWN *

[and REFUELINf ACTION: it o RPCW C i P I s t 1

or o T

a in a es or at ea oe 1"5

CO P E O e p 1 er O g v vin 0 E I $ -+ ^1 ~ -

n o r t g ve ea t ity c ngg. / \ ' [ Y SURVEILLANCE REQUIREMENTS-. -

4.6.2.2 No additional surveillance requirteents other than those identifled per Specification 4.6.2.1.

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

[ Th v c es e ut h e on s

r in in js ,ay s a E a o E s

" s n e a o PS nud s T

f r* * ; d c "N ?

$ Specification 3.0 @ (provides determine the methodology and necessary data to the appropriate time interval to reach a CALCULATED BULK CORE TEMPERATURE of 760 degrees F, t-U M LNFO ONLY

L c o 3. 4, 2. L Ins erT A 4

The Reactor Plant Cooling Water (RPCW)/PCR'V Liner Cooling System (LCS) shall be OPERABLE with;

a. One RPCW/PCRV LCS loop OPERATING with at least one heat exchanger and one pump in each loop OPERATING. and

b. With firewater supply available via one OPERABLE flow path.

Inserr G With no RPCW/PCRV LCS loop OPERATING, within /0 minutes be in at least SHUTDOWN and suspend all operations involving CORE ALT 4 RATIONS, control rod movements resulting in positive reactivity chang',s, or movement of IRRADIATED FUEL, and:

a. Restore at least one loop to OPE TING status prior to reaching a CALCULATED BULK CORE TEMPERATURE of 760 degrees F, or

b. Restore forced circulation cooling prior to reaching a CALCUIATED BUlX CORE TEMPERATURE 5of 760 degrees F.

Amendren8 No.

Page 3/4 6-DRAFT \

The use of 760 dagrees F CALCULATED BULK CORE FFD e e PEMPEllAfuld6s a division between the APPLICABILITY of Specification 3.6.2.1 3.6.2.2 is explained as follows: and In the FSV HTGR, inlet t:mperature greater than 760the limiting parameter of interest is a core f$C a BULK CORE TEMPERATURE degrees is acore F. The CALCULATED conservative calculation of the h C 645 maximum components.

potential temperature in the and surrounding The conservatism are such that if the CALCULATED I* BULK CORE TEMPERATURE is limited to 760 degrees F, the design inlet temperature of 760 degrees F is not exceeded. Systems used d f' U ~ for accident prevention and mitigation are required to satisfy refercict the single failure criterion whenever CALCULATED BULK CORE TEMPERATURE ts greater than 760 degrees F. However, when' CALCULATED BULK CORE TEMPERATURE is equal to or less than 760 gfCg g,. g degrees F, it is acceptable to require only one OPERABLE system 3*O'g* required for accident prevention and mitigation as acceptable without single core cooling failure consideration, on the basis of the limited 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 recalcu14 tion 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 taken that provided out forced of service for maintenance circulation is maintained. or surveillance The time to reach CALCULATED BULK CORE TEMPERATURE equal to 760 F degrees may be recalculated as often as required.

j

i Amendment No.

Page 3/4.7-DRAFT j PLANT AND SAFE SHUTDOWN COOLING SUPPORT FEB 2SYSTEMS 8 G33 3/4.7.1 TURBINE CYCLE ,

SAFETY VALVES - SHUT 00WN 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.

APPLICABit.!TY: SHUT 00WN and REFUELING ACTION: With less than the above required safety v41ve(s) OPERABLE, restore the required safety valve (s) to OPERABJ status prior ,

F or suspend all operationsto reaching a CALCULATED BUL involving CORE ALTERATIONS or con trol rod movements resulting in positive reactivity changes.

SURVE1LLANCE REQUIREMENTS._

4.7.1.6 No additional surveillance per Specification 4.7.1.5. required beyond those identified p5 c need , T . p ro v

d e * ['

Y Specification 3.0@provides the methodology and necessary data to determine the appropriate time interval to reach a CALCULATED BUIX CORE TEMPERATURE of 760 degrees F.

FOR

!NFO l

ON!.V l

l J

'. Amendment No.

Page 3/4.7-

. DRAFT BASIS FOR FEB 2 81986 SPECIFICATIONS LC0 3.7.1.5/SR 4.7.1.5 ANO t.C0

3. 7.1. 6/SR 4. 7.1. 6 The economizer-evaporator superheater (EES) section of eacn steam generator loop is protected by three spring-loaded safety valves, each with one-third nominal relieving capacity of each loop. The reheater section of each steam generator loop is protected from overpressure transients by a single safety valve. These steam generator 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, maintenance.

IGV requirements) every 5 years or after To satisfy the testing criteria, the valves must be tested with steam. Since these valves are-permanently installed j in . steam piping, the appropriate means for testing require plant power to be in excess of 22% RATED THERMAL POWER. Thus the test-must be conducted during LOW POWER. Conditions are spe,cified so as to minlatze operation at power until the valves are tested.

Due to the infrequent required testing of these valves, the likelihood of an accident oqcurring without proper valve

'a considered very small and plant safety is not compromised. testing During all MODES, with one EES safety valve inoperable, plant operation is restricted to a condition for which the remaining l sa fe ty valves have sufficient relieving capability to prevent {

I ove' pressurization of any steam genera tor section (i.e., one boiler feed pump per operating loop). Conversely, with any k reheater safety valve inoperable, plant operation is restricted l i

. to a more restrictive Mode.

A 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action time for repair or SHUTOOWN due to inoperable safety in valves ensures that .these valves are returned to se rvice a relatively short period of time, during which an overpressure h" ars d o t ". nottransient is unlikely. Operation at pewer for 72 tv any exts Aded period.result in a significant loss of safety function The setpoints for the safety valves identified in Table 4.7.1-1 are those valves identified in the FSAR with tolerances applied such that the Technical Specifications incorporate an upper bound setpoint. This is consistent with ng incorporating normal operating limits in these Specifications.

7 g c n e e.ej s To provide. t d io n e l e for C ALC vt AT E D 3 U L k:. e_ o R E TEMPER-Os.ss3 .

ATvRE. 3w2 Io c ro rt - r ef e re n c e 4 pee { 4 ; c c T Io " 7 . 0, N .

~encment No.

+ Page 3/4 7 21 1 l

PLANT AND SAFE SHUTOCWN COOLING SUPPORT SYSTEMS DRAFT !

i 3/4.7.4 SERVICE WATER SYSTEM-SHUTOCWN NOV 3 885 LIMITING CONDITION FOR OPERATICN -

l 3.7.4.2 The service water system shall be OPERABLE with:

a.

{

One- service 1 OPERABLE, water pump (P-4201, P-4202, or P 42025) I

b. An of OPERABLE service flow path to SAFE SHUTOOWN COOLING users water emergency diesel coolers, instrument air compres(sor and after coolers, and the reactor plant cooling water /PCRV liner cooling heat exchangers; E-4601, E-4602, E-4603, and E-4604), and

c. .

An OPERABLE flow path from the circulating water makeup system to the service water pump pit.

Appl!CABILITY:

STARTUP", SHUTDOWN *, and REFUELING ACTION:' With any of the requirements identified in 3.7.4.2 4, b,

.(.'

i or c above inoperable, restore the e:uipment to CPERABLE status orge to reaching

! TEMPERATURFof 760 degrees F, andasuscendCALCULATED BULK CORE all operations involving CORE ALTERATIONS or control rod movements resulting in positive reactivity changes.

l I

With

I'

the CALCULATED SULK CORE TEMPERATURE less than or equal to 760 t

,/ P S C seed To p r o v ,* d e % f) G o

fh Specification 3.0($fprovides determine the methodology and necessary data to TEMPERATURE of 760 degrees F.the appropriate time interval to reach al 3

W

_ , _,, _w~--- - - - - " - ' ' -

URM i Qe"'$'[.6 HOV 3 0 was During POWER, BULK CORE TEMPERATURE greater thanLOW POWER, STARTUP, and SHUT 760 degrees F, with only on, service water pump OPERABLE, a restoratt'en time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is provided to restore another pump to CPERABLE status. During this 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , service water needs can be met by the redundant pump.

With an inoperable flow path to the emer instrument air compressors and af ter coolers, gency diesel coolers or cooling is required, initiation of backuo because the affected component may be automatically initiated tp perform SAFE SHUT 00WN COOLING.

inoperable flow path With an cooling heat exchangers, backup cooling capability must beto the reactor p verified, i so that it can be manually initiated if recured. Actual initiation i

of backup cooling to the PCRV LCS heat exchangers is undesirable except in an actual fouling. Seventy-two hours emergency to minimize pos sibility of tube OPERABLE is provided to restore the flow path to status, this is provided for in the system consistent with OPERABILITY the restoration times requirements components. If the flow path cannot be restored within for the above backup cooling will be initiated 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

verified within (firewater) or the capability Both firewater pumps must be OPERABLE.1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to restore the equipment t for The Surveillance Requirements firewater pump OPERABILITY are given in Specification 3.5.4 If neither within 24 of these hours.conditions can be met, the plant must be in SHUTOOWN j If only one firewater pump is OPERABLE, a second t pump must be restored to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or the plant must {

be in SHUT 00WN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . {

( The 1 h ur restoration time is '

required because service water is unavailable, and the firewater supply has no redundant capability. If e firewater pumps are OPERABLE, the plant must be in SHUT 00WN witnin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months because service water and backup cooling (firewater) are both unavailable.

With the flow path from the circulating water makeup system to the service water pump pit (noperable, a restoration time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is

provided to restore the flow path to OPERABLE status. This ACTION gc time is adequate, since backup cooling (firewater) can be and considering the makeup requirements of the service water system. initiated, wendt During STARTUP, SHUT 00VN, and 7* [ CORE TEMPERATURE pump, less than or equal to 7 REFUELING, with the CALCULATED BULK cens- a flow path frem the circulating water makeuo system toflow p M g .,q the service water pump pit are required to be OPERABLE prior to the time calculated to reach a CALCULATED BULK CORE TEMPERATURE of 760 d F.

b f y ,.,gt;,, Thiswhen condition ACTION ensures that the plant will remain in a stable service 3 , 9, p) , esting, or unanticipated'water is unavailable due to maintenance, outages.

No specific surveillance are required as the service water system is normally flow pathoperating, including the required OPERABLE flow paths.The by the surveillance testing of the diesel generators.to the stancby die

~

. ENCLOSURE 5 Categorization of the Auxiliary Electrical Power Systems, 3/4.8, connents as transmitted to PSC by the NRC Letter of May 6,1987, Enclosures .1 and 2. Categortration symbols are the same as in previous meetings and correspondence (for example, the C. Hinson memorandum of j December 15, 1986).

l i

1

i l.

I' ,

e o

l Comment No. _,

Categorization 'l 1

ENCLOSURE i .l 1 -

0 2 0 3 D 4 0 5 0 6 0 7 0 ENCLOSURE 2 1 8 ,

2 D 3 'D l da 8 4b 8-4c D

! 4d 8 I

4e 8 5 D 6 0 7 D*

8 D ;

9 0 !

10 0 11 D ,

12 0 j 13 D 14a F 14b,c D 15 D 16 0 for reactivity

, change portion 17 0 18 0 19 D 20 -D 21 0 22 0 23 D 24 0 1 25 0-26 0 27 0 28a D*

28b 8 28c D*

2 1

, . 0 1

1 3

I l

Ccmment No. Catecoriration ENCLOSURE 2 (Continued) 28d B 28e B 28f A# DA# for i Licensee action l to address 5 days j vs 2 hrs.

28g 0 29h B

29) A# l

29) B l

l l

l I

l l

1 3

h I 5 1 40

/pn RE0u9'o,,

UNITED STATES e l' r, NUCLEAR REGULATORY COMMISSION 3 - ,E WASHINGTON D.C.20555 ~

July 2,1987 (

g*..../

Docket No. 50-267 l

Mr. R. O. Williams, Jr. !

1 Vice President, Nuclear Operations Public Service Company of Colorado Post Office Box 840 Denver, Colorado 80201-0840

Dear Mr. Williams:

l j

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR TECHNICAL SPECIFICATION UPGRADE PROGRAM FOR FORT ST. VRAIN NUCLEAR GENERATING STATION .

l (TAC NO. 56565) l 1

References:

See Enclosure 1 ,

l We have reviewed the information you submitted with your letter, Reference 1, on the Technical Specification Upgrade Program for the Fort St. Vrain Nuclear Generating Station (FSV). Your submittal, provided documentation of comment resolutions that were reached between Public Service Company of Colorado (PSC) and the Nuclear Regulatory Commission (NRC) during a meeting on October 27-30, 1986, regarding the FSV Technical Specification Upgrade Program (TSUP). This I 1

meeting was held to resolve comments provided by the NRC in Reference 2, on the November 30, 1985 FSV TSUP draft submittal. The NRC provided a meeting summary in Reference 3, which tabulates the comments discussed and the nature of their J l

resolution.

This letter provides a request for additional information (See Enclosure 2) as a result of our review of your current submittal, against our original comments (Reference 2). The comments are identified in the same manner as used in our meeting summary (Reference 3). Also, in Enclosure 2 we have provided requested actions for PSC resulting from our review and resolution of the eleven NRC action items from the meeting of October 27-30, 1986, and for the action items summarized in Enclosure 1 to Reference 2 that still require further PSC action.

Further, Enclosures 3 and 4 contain supplements to our comments transmitted in !

Reference 4 on the safety-related cooling functions. These supplemental l comments address our view of how your proposed revision to Technical Specifica-tion LCOs 4.3.1 (Reference 6) and 4.1.9 (Reference 7) impacts the review of the final draft of the FSV TSUP. This is in response to your desire to have the reviewers of the TSUP final draft satisfied that your proposed LCO 4.1.9 resolves any previous comments or concerns raised during the TSUP final draft review. Enclosure 3 provides the supplemental concerns and their proposed resolutions. Enclosure 4 is our markup of the November 30, 1985 draft Technical Specifications that incorporate our proposed resolutions.

y % .)

- -3 , a j i i Ji I

l

, o o Enclosure 5 transmits our proposed categorization of our comments on the Auxiliary Electrical Power Systems, Section 3/4.8, as transmitted to you in Reference 5. This proposed categorization uses the same designations as used in previous correspondence, for example, Reference 3. This proposed categoriza-tion is for your information and use in preparing your response to these comments.

Also, PSC should consider transmitting in TSUP format all Fort St. Vrain Technical Specification Amendments that have been issued in parallel with, but l not included in the TSUP final draft of November 30, 1985, and any that may be issued in the future, but prior to the TSUP amendment application. As a l minimum this would include Amendment Nos. 47 through 55. Receipt of the l material in the TSUP format will expedite our review and transmittal of any comments prior to this material being submitted in the TSUP amendment l application.

I Finally, our review of your TSUP amendment application will be expedited !

further if you would transmit your proposed resolutions to the forty-two "A#" l (PSC action items) and twenty-four "C" (PSC to explain in proposed safety (

evaluation) Category items (See the list of these items in the October 27-30, j l 1986 meeting summary, Reference 4). Again, this will allow us the opportunity j l to review and comment on these resolutions prior to their being submitted in the TSUP amendment application. ,

Please provide the required information within 60 days of receipt of this letter. If you feel that further discussion would be helpful in resolving !

these open issues, please call me at (301) 492-7592. l

)

The information requested in this letter affects fewer than 10 respondents; therefore, OMB clearance is not required under P.L.96-511.

Sincerely, l$

Kenneth L. Heitner, Project Manager !

Project Directorate - IV Division of Reactor Projects - III, IV, V and Special Projects !

Office of Nuclear Reactor Regulation l

Enclosures:

l As stated cc w/ enclosures:

See next page DISTRIBUTION Docket File NRC PDR L PDR E.6Ahv PD4 Reading D. Crutchfield . Schroeder J. CalvoP. Noonan K. Heitner OGC-Bethesda E. Jordan J. Partlow ACRS (10)

J. Miller , R. Emch PD4 P ant File g

PD4/LA PNoonan N

h PD4/PM KHeitner: sr h PD4/D JCalvo er TSB REmc 7/l/87 7/)/87 7/v/87 /1/87 7/J /87

Mr. R. O. Williams Public Service Company of Colorado Fort St. Vrain ,

CC:

Mr. D. W. Warembourg, Manager Albert J. Hazle, Director Nuclear Engineering Division Radiation Control Division Public Service Company Department of Health of Colorado 4210 East lith 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 191/2 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 Public Service Company of Colorado Denver, Colorado 80201 16805 Weld County Road 19-1/2 Platteville, Colorado. 80651 Senior Resident Inspector I U.S. Nuclear Regulatory Commission Mr. R. F. Walker i P. 0. Box 840 Public Service Company of Colorado Platteville, Cnlorado 80651 Post Office Box 840 Denver, Colorado 80201-0840 '

Kelley, Stansfield & 0'Donnell Public Service Company Building Commitment Control Program Room 900 Coordinator 550 15th Street Public Service Company of Colorado Denver, Colorado 80202 2420 W. 26th Ave. Suite 100-0 Denver, Colorado 80211 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Chairman, Board of County Commissioners of Weld County, Colorado Greeley, Colorado 80631 Regional Representative Radiation Programs Environmental Protection Agency 1 Denver Place 999 18th Street, Suite 1300 Denver, Colorado 80202-2413

1 .

ENCLOSURE 1 REFERENCES

1. H. L. Brey letter to H. N. Berkow, Technical Specification Upgrade Program, February 20, 1987, Public Service Company of Colorado, (P-87063).

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

3. Memorandum for H._N. Berkow from C. S. Hinson, Summary of October 27-30, 1986 Meeting at Fort St. Vrain (FSV) to Discuss Staff Comments on the FSV Technical Specification Upgrade Program (TSUP), December'15, 1986, U.S.

Nuclear Regulatory Commission.

4. K. L. Heitner letter to R. O. Williams, Jr. , NRC Comments on the Technical Specification Upgrade Program (TSUP)' LCOs for' Safety-Related Cooling Function, April 17, 1987, U.S. Nuclear Regulatory Commission.

5. K. L. Heitner letter to R. 0. Williams, Jr. , Draft Updated Technical Specification 3/4.8, Request for Additional Information (TAC No.'56565),

May 6, 1987, U.S. Nuclear Regulatory Commission.

6. R. O. Williams letter to H. N. Berkow, Proposed Technical Specification Change Eliminating Reliance on Reheater Section of' Steam Generator for Safe Shutdown Cooling, January 15, 1987, Public Service Company of Colorado, (P-87002).

7. Draft R. O. Williams letter to H. N. Berkow, Proposed Technical Speciff-cation Change, LCO 4.1.9, Core Inlet Orifice Valves / Minimum Helium Flow and Maximum Core Region Temperature Rise, April 17, 1987, Public Service ~

Company of Colorado, (P-87124).

- ENCLOSURE 2 ADDITIONAL INFORMATION NEEDED TO COMPLETE REVIEW Of TECHNICAL SPECIFICATION UPGRADE PROGRAM FOR {

l FORT ST. VRAIN NUCLEAR GENERATING STATION l I

The following is a list of requests for additional information needed to complete this review. The identification of each item is the same as in the NRC notes of the October 27-30, 1986 meeting (G. L. Plumlee Memorandum {

Dated, December 1, 1986) unless noted otherwise. l

\

)

l l

TABLE 1.0-1 l l The Licensee should provide additional justification for their l proposed mark-up of Table 1.1 (P. 1-9, Attachment 1 of PSC letter of February 20, 1987) or should make the '@" and "#" footnotes specific to the precise evolutions (tests) or surveillance involved. PSC references such ;

provisions in the GE-BWR STS. However, neither the GE-BWR STS, NUREG-0123, l l

i Rev. 3, nor the Perry TS (example given out in October 27-30, 1986 meeting f i by PSC) allow indiscriminate switching as would be allowed with the PSC proposed words of "...for the purpose of performing surveillance or other tests...". Both the GE-BWR STS and the Perry TS allow switching of the Mode Switch Position only under rather precisely defined conditions, such as " . . .to test the switch interlock f unctions. . ." or ". . .while a single I

control rod drive is being removed from the reactor pressure vessel per Specification 3.9.10.1. . ." or ". . .while a single control rod is being recoupled or withdrawn provided that the one-rod-out interlock is OPERABLE".

l further, please revise the pound sign (#) footnote in Table 1.1 by deleting the words "provided that k eff is verified less than 0.99" and l

replacing the deleted words with "provided that the control rods are verified to remain fully inserted." As acknowledged in the PSC response to NRC Comment No. 4 on TSUP draft LC0 3.1.3, the use of the word " verified" is incorrect in establishing a quantitative estimate of SHUTDOWN MARGIN (that is, kg ) because the reactor operators cannot actually verify the accuracy of the calculated assessments provided to them. However, l

1

. o

- consistent with the equivalent footnote provided in Table 1.2 of the BWR-STS (NUREG-0123, Rev. 3), the operators can verify that control rods are fully inserted.

SL 2.1.1-5 The Licensee should provide additional information and a safety evaluation to support splitting the existing FSV Safety Limit 3.1, part l

into the proposed Safety Limit Section 2.1.1 (November 1985 Draft) and part into a Limiting Condition for Operation LCO 3.2.6 (November 1985 Draf t).

The 24 hr action time has been adequately addressed in PSC's letter of February 20, 1987. However, the reference to FSAR Revision 4 Section 3.6.8 does not provide specific justification for downgrading part of the existing Safety Limit to a Limiting Condition for Operation. In fact, FSAR Revision 4 Section 3.6.8 titled " Core Safety Limit" discusses all of the limits in the proposed SL 2.1.1 and LCO 3.2.6 as if they were safety limits as does the existing FSV SL 3.1.

LCO 3.0 - New Item

)

Please provide an equivalent LCO to the recently proposed LCO 4.0.4 in the existing Technical Specifications.

LC0 3.1.1-1 Please revise the wording of TSUP draft LC0 3.1.1.a to add the words "from the fully withdrawn position" immediately after the words "152 seconds." PSC was to propose an alternative to the May 30,1986, NRC narkup of the TSUP draf t. The proposed revisions to the basis does not adequately clarify the limiting condition for operation. In addition, the wording of the proposed insert at the bottom of page 3/4 1-5 in the PSC i

narkup (Attachment 1 to P-87063) needs to be revised to read as follows:

The full insertion scram time can be determined either directly from a full insertion scram time or indirectly from a partial scram time of 10 inches or more. For the partial scram time, the estimate of an 2

extrapolated scram time is always based on assuming scram from the fully withdrawn position and not from the actual rod position.

LC0 3.1.3-4,-8 Please provide additional information with regard to the PSC position on the substantiation and verification of nuclear methods as expressed in the PSC response to NRC Comment Nos. 4 and 8 that are given in Attachment 2 to P-87063. Is it PSC's position that the verification and documentation 1 of the assessment methodology used in LCO 3.1.3 meet the intent, if not all the specific requirements or guidelines, of industry standards per ANSI /ASME N45.2.11, ANSI /ANS-19.3 (Section 6), ANSI /ANS-19.4, ANSI /ANS-19.5, and ANSI /ANS-19.6.1? Are detailed records or other supporting documentation for the PSC position maintained from the reviews of the base reactivity curve as required to be performed by the Nuclear facility Safety Committee (NFSC) per the existing Technical Specifications LCO 4.1.8 and SR 5.1.47 Can it be demonstrated from the NFSC records that the type of analytical-to-experimental comparisons cited in the PSC position have been and are being factored into the NFSC review? If not, explain how the NFSC has verified the continued use of the base reactivity curve to assure an adequate SHUTDOWN MARGIN per the basis of existing LC0 4.1.8 without documenting and updating such comparisons as those cited in the PSC position? If such documentation is not being maintained, proposed TSUP draft Specification AC 6.5.2.7.1 needs to be changed such that the NFSC is specifically charged with reviewing and documenting the results of the performance of SR 4.1.3, SR 4.1.4.1.2, SR 4.1.5, and SR 4.1.7 and the results of comparing the reload design against the requirements specified in DESIGN FEATURE 5.3.4 The reviews and documentation are both necessary and appropriate so that there is a record of the verification process for assuring the adequacy of the SHUTDOWN MARGIN assessment methodology employed in LCO 3.1.3 and SR 4.1.3. This approach is consistent with the intended role of the NFSC as agreed upon with NRC and acknowledged by PSC in the Attachment to PSC letter, C. K. Millen to R. S. Boyd, September 11, 1975.

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- LC0 3.1.6 New Item (Old LC0 3.9.1-8)

The Licensee should add SHUTDOWN and REFUELING to the Applicability statement of LC031.6 on the Reserve Shutdown System. Such added i

applicability should account for the exception required for any two control rod pairs which may be removed from the PCRV (LC0 3.1.4.2.a.) of November 30, 1985 Draft). Although the May 30, 1986 NRC comment LCO 3.9.1-8 on inadequate SHUTDOWN MARGIN du ring REFUELING was categorized as an "F" (futher discussion possible) in the October 1-2, 1986 meeting, further comparison to the GE-BWR STS Rev. 3, NUREG-0123, indicate that acceptance of small SHUTDOWN MARGINS (LC0 3.1.1 P 3/41-1) during REFUELING with any control rod withdrawn is,with the proviso that the Standby Liquid Control System be OPERABLE (LC0 3.1.5, P 3/41-19). The GE-BWR STS is used ]'

for comparison rather than the W-STS as the Ft. St. Vrain reactivity control system of control rods and reserve shutdown material is similar to the GE reactivity control system of control rods and Standby Liquid Control System, with neither having routine boration reactivity control as in the Westinghouse system. Also, for Ft. St. Vrain, the Actions b.2 for SHUTDOWN ar:d C.2.b for REFUELING with inadequate SHUTDOWN MARGIN (LCO 3.1.3 of November 30, 1985 Draft) require actuation of sufficient reserve shutdown material to achieve the required SHUTDOWN MARGIN. Yet in the Reserve Shutdow. System Specification (LC0 3.1.6 of the November 30, 1965 draft) l there is no requirement for OPERABILITY in either the SHUTDOWN or REFUELING condition.

LC0 3.2.1-1 ,

4 Please expand the Basis to describe briefly what activity is involved in the " determination by evaluation." What reactor observables are evaluated?

LCO 3.3.2.3-3 i

The Licensee should add a calibration requirement in SR 4.3.2.3.2 for seismic instruments determined to be out of calibration following a seismic event. As indicated in the basis, P. 3/4 3-83 of the November 30, 1985 4

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- Draf t, the Licensee has already comitted to this (last sentence, fourth j paragraph). However, there is presently no requirement for such calibration in the surveillance. Although the STS guidance is to do such l calibration within ten days, it is judged that 30 days provides for i allowing monitoring of af ter shocks follwing the initial event. j Experience with other comercial plants is that the contractor / manufacturer !

provides on-site calibration capability. PSC should investigate such on-site service capability from their seismic instrument manufacturer so that the seismic instrun.ents remain on-site to be maximum 1y available for monitoring after shocks. i LC0 3.3.2.3-4 ,

i i l The Licensee should specify monthly CHANNEL CHECKS for the seismic instruments in TS Table 4.3.2-2 (November 1985 Draft, P. 3/4 3-82) rather j than the proposed quarterly checks. PSC had proposed quarterly CHANNEL CHECKS as consistent with Turkey Point seismic instrument surveillance.

However, as Turkey Point specifications are the exception to the rule and !

as Turkey Point is situated in Seismic Zone 0 and FSV is situated in ,

Seismic Zone 1, the Turkey Point specifications are not a reasonable comparable to use. Additionally, it is in the Licensee's interest to have ;

the seismic instruments operable to facilitate restart if shutdown occurred. The relatively uncomplicated CHANNEL CHECKS would enhance instrument operability and thus the assessment of a seismic disturbance and the implications to plant restart capability.

LCO 3.5.4 - New Item Please revise LC0 3.5.4 ACTION b (second part on page 3/4.5-30 of the TSUP draft). The words, " establish a backup system for fire suppression purposes within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ," need to be changed to the wording, " establish a backup system for fire suppression purposes prior to reaching a CALCULATED BtJLK CORE TEMPERATURE of 760"f but within a period of time not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ." The revision is necessary to assure an operable flow path to the liner cooling system (LCS) via the firewater system when the LCS is the only decay heat removal path during an interruption of forced cooling for 5

. purposes of maintenance and inspection. This change is consistent with the NRC guidelines provided in the NRC letttr of December 5, 1986, on PSC i commitments required to approve the proposed version to LCO 4.1.9 in the existing Technical Specifications, j

LC0 3.6.4-2 ]

Please provide additional.information with regard to the source and approval of the acceptance criteria for PCRV concrete permeability and PCRV liner thinning as cited in the PSC markup of pages 3/4.6-39 and 3/4.6-40 of the TSUP draft in Attachment 1 to P-87063. The NRC does not have a copy of the previous ISI procedures alluded to in the citations, and these acceptance criteria are not found in the recent ISIT submittals.

LC0 3.7.2-4 The Licensee should provide additional clarification in the basis on P. 3/4 7-14, third paragraph, second and third sentences, as they read:

"A 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ACTION time is provided to isolate the affected loop, in an effort to regain OPERABILITY of a second hydraulic fluid pump and/or at least one accumulator for the affected valve group. If OPERABILITY of a second hydraulic fluid pump and/or at least one accumulator is not restored within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , reactor shutdown is required with 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ."

Since OPERABILITY of a second hydraulic fluid pump is not mentioned in the Action Statement, only " supply of at least 2500 psig", the basis should use the terminology of the Action Statement, namely, supply of at least 2500 psig. Also, the basis has interpreted the Action Statement to' allow one hour to restore the required conditions of at least one accumulator and or at least 2500 psig pressure before reactor shutdown within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The Action Statement a. of P. 3/4 7-12 does not address a restoration time. It requires isolation of the affected loop in one hour and reactor shutdown in the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months without regard to any restoration time.

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LC0 3.7.6.3-3 For the halon system in Building 10, please propose specifications consistent with SR 4.7.6.3 in the WNP No. 2 Tec.hnical Specifications.

NUREG-1009 (see attached page). At WNP No. 2, tank " quantity" is determined once per six months using the heat tape and gun method approved' by the American Nuclear Insurers. However, storage tank weight must be ;

verified periodically. PSC needs to specify an appropriate surveillance period for verifying storage tank weight and to justify any period exceeding 36 months.

LCO 3.7.8-4 Please correct the misspellings in PSC's markup of TSUP draf t LC0 3.7.8 ACTION b. The word, " values," is misspelled twice as " valves."

LOC 3.7.10-3 Please revise FSAR Sections 1.4 and B.5.2.7 to be consistent with the Basis definition of " safety-related" as including Class la components. PSC expressed the desire to retain the wording in the opening sentence of the Basis as being indicative of their position with regard to the scope of the safety-related snubbers. The Basis and the FSAR need to be consistent on this point.

LCO 3.9.1-2 The Licensee should retain the APPLICABILITY as it was in the November 1985 Draf t or should provide additional information or chenges to split out Specifications 3.9.la and b from Specifications 3.9.lc and d.

Specifications 3.9.la and b are justly applicable to only "whenever both primary and secondary PCRV closures of any PCRV penetrations are removed" which is PSC's proposed APPLICABILITY. However, Specifications 3.9.1 c and d on requiring two startup channel neutron flux monitors and maintaining the SHUTDOWN MARGIN requirements of Specification 3.1.3, are 7

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PLANT SYSTDt$

HALON SYSTEMS ,

( t!NITING CON 0! TION FOR OPERATION i I

)

i 3.7.6.3 The 18 Halon systees in the PGCC units in the control 'roon j JMil be OP[RA8LE with the storage tanks having at least 95% of full charge weight and 905 of full charge pressure.

APPLICA8!tiTY: Wnever equipment protected by the Halon systees is required to be OPERA 8LE, ACTION:

a. With one or more of the above required Halon systees inoperable. I within 1 hear establish a continuous fire watch with backg fire l suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, estab11sh an hovely fire watch patrol. i I

b. The provf sfons of SpectfIcatiens 3.0.3 and 3.0.4 are not app 1fcable.

SURVEILLANCE RE001 REM [NT5 4.7.6.3 Each of the above required Halon systems shall be demonstrated .

OPERA 8LE: ,

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a. At least once per 31 days by vertfying that each valve (manual, I power-operated, or automatic) in the flow path is in its correct position.

b. At least once per 6 months by verifying Naton storage task quantity and pressure.

c. At least once per 18 months by:

1. . Verifying the system, includlag associated ventilation systee i fire dampers and fire door release mechantsas, actuates, j manually and automatically, upon receipt of a slaulated actuation sfgnal, and l

2. Forformance of a flow test through accessible headers and nonles to assure no blockage.

d. At least once per 36 months by vertfying Halon storage tank welght.

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WASHINGTON KKLEAR - UNIT 2 3/4 7 23 :

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1 applicable throughout the REFUELING mode. Therefore, either the original I APPLICABILITY in the November 1985 Draf t should be retained or 3.9.la and b .

I need separated from 3.9.1c and d with different applicability statements as j discussed above.

LCO 3.9.1-3 l

1 The Licensee should revise the wording as suggested for Action b of

!.00 3.9.1 (similar wording should also be used for Action C.1 of this LCO) l to: 1 1

I "With one of the above required neutron flux monitors inoperable, or not OPERATING, immediately suspend all operations involving CORE ALTERATIONS, any evolution resulting in positive reactivity changes, or movement of IRRADIATED FUEL."

l The Licensee's position that the proposed woros ". . . control rod movaments resulting in positive reacitvity changes . . ." were intended to be specific versus the NRC recommended wording ". . . any evolutions resulting in positive reactivity changes . . ." has been accepted for the other LCOs involved LCO's 3.6.5.1-2, 3.6.5.2-1, 3.7.1.1-2, and 3.7.9-3 (other NRC actions items from October 27-30, 1986 meeting concerning reactivity changes). For those LCO's, flexibility to change between fecdwater and condensate cooldown is accepted because the positive reactivity effect of any involved cooldown has already been accounted for in the SHUTDOWN MAR 61N requirement of 0.01 delta k in accordance with LC0 3.1.3. And the reactivity additions are relatively small and added f

slowly as PSC explained in Attachment 4 to their February 20, 1987 letter, for the subject LCO, however, Actions b and and C.1 involve loss of one and both startup channel neutron flux monitors, respectively. Allowing an intended positive reactivity change due to cooldown with degraded startup l channel neutron flux monitoring is unacceptable. When both startup channels are inoperable, any controllable positive reactivity change should be stopped, as under these conditions, immediate assessment of flux changes is lost. With only one operable startup channel of neutron flux monitoring, it is also unacceptable to intentionally make positive l

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reactivity changes as that one startup channel may be operating erroneously and there is no second operating channel to confirm it's readings. Also, with only one operating channel, a sudden 1 css cf that channel again results in complete loss of the ability to make immediate assessments of neutron flux changes. These changes would make the FSV Actions consistent with those of STS Rev. 5, P. 3/4 9-2, on neutron flux monitoring capability during refueling, and with those of the existing FSV Technical Specifications, LC0 4.7.1.

AC 6.3/6.4-1 The Licensee should include reference to the NRC March 28, 1980 letter l

in Technical Specification Sections 6.3.2 and 6.4.1 per the subject NRC c ommen t . Contrary te the Licensee's position in the October 27-30, 1986 Meeting that the NRC March 28, 1980 letter is not in their letter log and therefore is doubtful that is was agreed to, the Licensee responded to it in their letter of December 20, 1980 (P-80438). In Attachment 1 of that letter, the Licensee stated that although they had received the NRC March 28, 1980 letter too late to implement the requirements by August 1, 1980 as required by the letter's Enclosure 1, some requirements would be met by the August 1,1980 date and their program for compliance would be submitted by January 15, 1981. As it appears that the Licensee has committed to the subject NRC March 28,1980 requirements, stating this in the Technical Specifications should not involve any undue hardships.

AC 6.5.1.6a (AC 6.5-1)

The Licensee should revise AC 6.5.1.6a to require the PORC to review any procedures required by AC 6.8.4 as AC 6.8.4 includes procedures for iodine sampling in the reactor building and Post-Accident Sampling, both of which are THI-2 Action items of NUREG-0737 and Generic Letters 83-36, 31.

Also, the Licensee should revise AC 6.5.2.7 to require the NFSC to review the programs of AC 6.8.4. This is what PSC said PORC and NFSC do now (See PSC letter of February 20, 1987, Attachment 2). However, as written, AC 6.5.1.6a and AC 6.5.2.7 do not have these requirements but should.

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. AC 6.5.1.7b (AC 6.5-1)

The Licensee should provide additional information.to explain what 1

" Procedure Deviation Reports" are, as used in their letter of February 20, 1987, Attachment 3, Item AC 6.5# 1-7 and Attachment 1, ;

P. 6-15, narked up item 6.5.1.7.b. In Attachment 3, reference is nede to l

' Temporary Changes" whereas in Attachment 2, reference is made to ;

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" Procedure Deviation Reports". It is not clear if these are one and the l same thing. If they are the same, PSC should provide additional information to justify the proposed Exception for " Temporary Changes".

AC 6.9.1.2.a (AC 6.9-1)

The Licensee should revise their proposed marked up P. 6-27 (PSC letter of February 20, 1987, Attachment 1) to either put report submittal times in 6.9.1.2, first paragraph as in the STS Rev. 5, P. 6-16, or place a repor t submittal directly in 6.9.1.2.a.2. As proposed, 6.9.1.2.a.2 does not have any report submittal time connected to it. PSC placed their report submittal directly in 6.9.1.2.a.1 rather than 6.9.1.2. Therefore, when 6.9.1.2.a.2 was added, it was not covered by a report submittal time.

The following RAI is a result of the review of Attachment 1 to P.-85448 NRC Action 3 PSC should provide the Environmental Qualification study for this item. l The following is NRC Action Item 3 and its response from Attachment 1 to PSC letter of November 27,1985 (P.- 85448). !

1 NRC Action Provide guidance on whether components, which are required to function i

to ma',ntain other equipment within an environment for which it is qualified, should also be in the Technical Specifications (for example, main steam isolation valves).

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- NRC Response Where assumptions for equipment operability related to environmental qualification are' based on the successful operation of active components in the event of an accident, the availability and reliability of these components should be ensured through Technical Specification requirements.

Specific items of concern identified by the staff were the main steam isolation valve (MSIV) operability and closing time requirements as well as the hot reheat (HRH) valve operability requirements. Other EQ operability I

requirements should be identified by the licensee and incorporated as Technical Specification requirements to ensure equipment necessary to mitigate accidents function within the assumed environmental conditions.

In addition to the EQ analysis, other analysis which rely on equipment operability may also result in Technical Specification requirements, for i example, the equipment operability requirements based on previous fire -l protection analysis (Appendix R Evaluation: FSV Reports 1 through 4) should be reflected in the Technical Specifications, i i

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ENCLOSURE S Additional Comments on The Fort St. Vrain Technical Specification Upgrade Program Final Draft LCOs for Safety Related Cooling Functions )

1 NRC COMMENTS-LCO 3.5.1.1

1. Based on the PSC letter (P-87002) dated January 15, 1987, the condition statement for LCO 3.5.1.1.a.2 needs to be rewritten as follows:

Both steam generator sections (both the economizer-evaporator-superheater (EES) and the reheater) OPERABLE including two OPERABLE flow paths.

2. Previously, the safe shutdown cooling outlet flow paths were via the by-pass valves off each loop's superheater outlet and each loop's hot reheat steam line. The by pass valves were verified to be OPERABLE as part of the normal operation of the bypass function. The recently installed six inch vent lines described in P-87002 are apparently not to be used on a routine basis. Therefore, SR 4.5.1.1.b needs to be revised by renum'oering surveillance b.2 and b.3 to b.3 and b,4, respectively, and adding a new ]

surveillance as SR 4.5.1.1.b.2. The new SR 4.5.1.1.b.2 should read as follows:

At least once per 18 months by verifying the OPERABILITY of each superheater outlet flow path by verifying that the valver..in the six-inch vent lines can be opened and that the vent flow pathr, are not obstructed.

3. Subject to NRC final approval of the proposed revisions to the Basis for the existing LCO 4.1.9, the following paragraph needs to be added at the bottom of the fourth page of the Basis for.LCO 3.5.1.1 following the second paragraph of the subsection entitled Redundancy Criteria:

Specification 3.0.N provides the methodology and necessary data to determine the appropriate time interval to reach a CALCUIATED BULK CORE TEMPERATURE of 760 degrees F. If the' active core remains below this temperature, which corresponds to the design maximum core inlet temperature as indicated above, then the design core inlet temperature cannot be exceeded and there can be no damage to fuel or PCRV internal components regardless of the amount, including total absence, or J reversal, of primary coolant helium flow. '

PSC needs to provide the appropriate number "N" for the cross referenced specification. The need for Specification 3.0.N has been identified in the NRC Request for Additional Information at Enclosure ?..

4. The first para 5raph of the subsection entitled Steam Cenerators on the fifth page of the Basis for LCO 3.5.1.1 and LCO 3.5.1.2 needs to be replaced with the following paragraph. 4

2 Whenever the CALCULATED BULK CORE TEMPERATURE exceeds 760 degrees F, both the reheater and EES sections of the steam generator must be OPERABLE. The steam generator reheater or EES sections can receive water from either the emergency condensate header or the emergency feedwater header as required to be OPERABLE per this Specification and per Specification 3/4.5.3. System flow OPERABILITY is determined by verifying flow from each of the aforementioned emergency headers (see i LCO 3.5.3.1) through each section of each steam generator. Whenever the CALCULATED BULK CORE TEMPERATURE is less than or equal to 760 degrees F or the plant OPERATIONAL MODE is REFUELING, system flow OPERABILITY is determined by verifying flow from either of the aforementioned emergency headers (see LCO 3.5.3.2) through either section of either steam generator. )

5. An additional paragraph is also required under the subsection entitled ,

Steam Generators on the fifth page of the Basis for LCO 3.5.1.1. This l paragraph needs to discuss the appropriate operability requirements for the '

seismically and environmentally qualified six inch vent lines and to cite j the supporting safety analysis requiring the use of these vent lines.

PROPOSED RESOLUTIONS 1./2./3./4./5. PSC needs to incorporate the required changes.

NRC COKMENTS-LCO 3.5.1.2

1. Contrary to NRC Comment No. 1 on LCO 3.5.1.2 as given in Enclosure 1 l to the NRC letter, Heitner to Williams, April 17, 1987, the words to be .

added after the word "0PERABLE" in the condition statement for LCO 3.5.1.2.a.2 need to be " including one OPERABLE flow path," not two. LCO 3.5.3.2 requires only one flow path, either the emergency condensate header or the emergency feedwater header, to be OPERABLE in STARTUP and SHUTDOWN whenever the CALCULATED BULK CORE TEMPERATURE is less than or equal to 760 degrees F or in REFUELING.

2. See NRC Comment No. 4 on LCO 3.5.1.1.

3. Delete LCO 3.5.1.2.b.1 and renumber b.2 through b.5 as b.1 through b.4, respectively. The deleted LCO is neither needed nor appropriate since the boiler feed pumps are not required to be OPERABLE under the same APPLICABILITY statement (see LCO 3.7.1.1) nor necessarily the emergency feedwater header (see LCO 3.5.3.2). For decay heat levels below that for shutdown from 35% power, previous analysis has shown that fuel damage will not occur in the depressurized core as long as one train of the PCRV LCS is operating. Necessary modifications to the TSUP draft to incorporate the guidelines from the NRC letter, Heitner to Williams, dated December 5, 1986, are indicated elsewhere in these comments, and these modifications are directed to providing assurance that the PCRV LCS provides a backup cooling capability during the conditions of APPLICABILITY. However, NRC reserves the right to modify LCOs 3.5.1.2, 3.5.3.2, and 3.7.1.1 pending completion of the review on the occurrence frequency of rapid depressurization events.

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, 3 In addition, further modifications to Specifications 3/4.5.1, 3/4.5.3, and 3/4.7.1.1 may be required depending upon the final resolution to NRC Comment No. 1 on Section 3/4,4 as documented in Enclosure 1. to the NRC letter, Heitner to Williams, April 17, 1987. As noted in the subject comment, the Updated FSAR does not support operation with the reactor producing fission heat simultaneously with reactor cooling using condensate only. When fission heat is being produced in the. critical core, normal cooling should be provided by forced circulation using either steam-drive or feedwater-drive of the helium circulators. The current wording of Specifications 3/4.5.3 and 3/4.7.1.1 do not provide for the OPERABILITY of the feedwater-drive for the safety-related emergency core cooling function when the CALCULATED BULK. CORE -TEMPERATURE is less than 760 degrees F although fission heat is allowed to be as high as 5% of rated reactor power in STARTUP. If new specifications are not added to address the OPERABILITY of the feedwater-drive for the_ important-to-safety normal cooling function ;

whenever fission heat- is being produced, further restrictions may be )

required.

One option is to delete the footnote on STARTUP in LCOs 3.5.1.1, 3.5.3.1, and 3.7.1.1 and to delete STARTUP in the APPLICABILITY in LCOs 3.5.1.2 and 3.5.3.2. Alternately, to allow the flexibility of performing training i starts while limiting the allowed fission heat level, the footnotes on the subject LCOs can be changed as follows. In LCOs 3.5.1.1, 3.5.3.1, and 3.7.1.1, the footnote should be changed to read: "Whenever CALCULATED BULK CORE TEMPERATURE is greater than 760 degrees F, or, in STARTUP, whenever ,

reactor thermal power is equal to or greater than 2 percent of rated reactor power." In LCOs 3.5.1.2 and 3.5.3.2, the footnotes should be changed to read: "Whenever CALCULATED BULK CORE TEMPERATURE is less than or equal to 760 degrees F, or, in STARTUP, whenever reactor thermal power is less than 2 percent of rated reactor power."_In addition, the Bases for the affected LCOs should be modified to indicate that the limits on STARTUP are to allow flexibility for achieving criticality during training starts and ~

that procedures have been implemented to limit the amount of fission heat to much less than 2% during such activity.

4. LCO 3.5.1.2 ACTION b needs to be rewritten as follows:

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b. With less than the above required OPERABLE equipment and with no forced circulation being maintained, be in at least SKUTDOWN within 10 minutes and suspend all operations involving CORE ALTERATIONS, control rod movements resulting in positive reactivity changes, or l movement of IRRADIATED FUEL, and either: l

1. Restore forced circulation on at least one loop prior to reaching a CALCULATED BULK CORE TEMPERATURE of 760 degrees F, and comply with ACTION a, or, j 1

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

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5. In LCO 3.5.1.2, both ACTIONS a and b, the words " CALCULATED BULK CORE TEMPERATURE" are used with regard to the required restoration of equipment or conditions. In both instances, a footnote symbol should be added after the word " TEMPERATURE" with'the following words provided in the text of the footnote:

Specification 3.0.N provides the methodology and necessary data to determine the appropriate time interval to reach a CALCULATED BU1K CORE TEMPERATURE of 760 degrees F.

PSC needs to provide the appropriate number "N" for the cross referenced specification.

1 PROPOSED RESOLUTIONS 1./2./3./4 /5. PSC needs to incorporate the required changes.

3. PSC needs to provide additional specifications per the previous cited comment on Section 3/4.4 or to incorporate the suggested changes in the LCOs cited or to propose and justify alternatives.

NRC C0KMENTS-LCO 3.5.3.1

1. In SR 4.5.3.1, the reference to Specifications 4.5.2.1.,a needs to be i deleted and replaced with a reference to Specification 4.5.1.1.b.1.

2. See above NRC Comment No. 3 on LCO 3.5.1.2. j I

PROPOSED RESOLUTIONS j i

1. PSC needs to incorporate the required changes.

2. Same as above for NRC Comment No. 3 on LCO 3.5.1.2.

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NRC C0KMENTS-LCO 3.5.3.2 l

1. The ACTION for LCO 3.5.3.2 needs to be deleted and replaced as follows:

With both the emergency feedwater and emergency condensats header inoperable, be in at least SHUTDOWN within 10 minutes and restore at least one header to OPERABLE status prior to reaching a CALCULATED BULK CORE TEMPERATURE of 760 degrees F and suspend all operations involving CORE ALTERATIONS, control rod movements resulting in positive reactivity changes, or movement of IRRADIATED FUEL.

2. In the ACTION, PSC should add the same footnote as described above in NRC Comment No. 5 on LCO 3.5.1.2. Similar information and cross references as provided in the footnote should also be included in the Basis.

3. In SR 4.5.3.2, the reference to Specification 4.5.2.1.a needs to be deleted and replaced with a reference to Specification 4.5.1.1.b.1.

4 See above NR.C Comment No. 3 on LCO 3.5.1.2.

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. 1 PROPOSED RESOLUTIONS 1./2./3. PSC needs to incorporate the required changes.

4. Same as above for NRC Comment No. 3 on LCO 3.5.1.2.

NRC COMMENT-LCO 3.5.4 In ACTION b (the second part of the ACTION statement on page 3/4 5-30), PSC should add the same footnote as described above in NRC Comment No. 5 on LCO 3.5.1.2. A request to modify this ACTION has been provided in the NRC Request for Additional Information at Enclosure 2. The words "within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months " need to be replaced with the words " prior to reaching a CALCULATED BULK CORE TEMPERATURE of 760 degrees F but within a period of time not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ."

Similar information and cross references as provided in the footnote should also be included in the Basis.

PROPOSED RESOLUTION PSC needs to incorporate the required changes.

NRC COKMENTS-LCO 3.6.2.2

1. In LCO 3.6.2.2, the condition statement should be deleted and replaced with the following:

The Reactor Plant Cooling Water (RPCW)/PCRV Liner Cooling System (LCS) shall be OPERABLE with:

a. One RPCW/PCRV LCS loop OPERATING with at least-one heat exchanger i and one pump in each loop OPERATING, and I

b. With firewater supply available via one OPERABLE flow path.

The change is necessitated to comply with the NRC guidelines for PSC commitments with regard to proposed revisions to existing LCO 4.1.9. These guidelines are given in the NRC letter, Heitner to Williams, dated December 5, 1986.

2. In LCO 3.6.2.2, the ACTION statement should be deleted and replaced with the following:

With no RPCW/PCRV LCS loop OPERATING, within 10 minutes, be in at' least SHUTDOWN and suspend all operations involving CORE ALTERATIONS, control rod movements resulting in positive reactivity changes, or movement of IRRADIATED FUEL, and:

a. Restore at least one loop to OPERATING' status prior to reaching a CALCULATED BULK CORE TEMPERATURE of 760 degrees F', or

b. Restore forced circulation cooling prior to reaching a CALCULATED BULK CORE TEMPERATURE of 760 degrees F.

6 1

The change is necessitated to comply with the NRC guidelines for PSC-commitments with regard to proposed revisions to existing LCO 4.1.9. These guidelines are given in the NRC letter, Heitner to -Williams, dated December 5, 1986. ,

i

3. If the interfacing isolation valves between the . firewater. system 'and .l the RPCW/PCRV LCS are not covered in the surveillance on the SAFE SHUTDOWN COOLING water supply system per SR 4.5.4.1.f or SR 4.5.4.1.g.3, the subject -

isolation valtes need to be covered by revising SR 4.6.2.2 appropriately.

4. In the ACTION, PSC should add the same footnote as described above in NRC Comment No. 5 on LCO 3.5.1.2 Similar information and cross references as provided in the footnote should also be included in the Basic.

PROPOSED RESOLUTIONS 1./2./3./4. PSC needs to incorporate the required changes.

NRC COMMENT-LCO 3.7.1.1 j See above NRC Comment No. 3 on LCO 3.5.1.2.

I PROPOSED RESOLUTION I Same as above for NRC Comment No. 3 on LCO 3.5.1.2. l l

NRC COMMENT-LCO 3.7.1.6 l In the ACTION, PSC should add the same footnote as described above in NRC-Comment No. 5 on LCO 3.5.1.2. Similar information and cross references as provided in the footnote should also be included inLthe Basis.

PROPOSED RESOLUTION PSC needs to incorporate the required changes.

NRC COMMENT-LCO 3.7.4.2 1

In the ACTION, PSC should add the same footnote as described above in NRC Comment No. 5 on LCO 3.5.1.2. Sicilar information and cross references as provided in the footnote should also be included in the Basis.

PROPOSED RESOLUTIONS I PSC needs to incorporate the required changes.

l 4

i i

.2 -w

I Amendment No. ENCLOSURE 4

'. Page 3/4.5-bi$/k,IY -

SAFE SHUT 00VN C0 CLING SYSTEMS 3/4.5.1 SAFE SHUT 00WN COOLING EQUIPMENT ,

LIMITING CON 0! TION FOR OPERATION 3.S .1.1. a . Two primary coolant loops shall be OPERABLE, each with at least: .

1.

Oneglium circulator OPERABLE, and

2. steam generatorsection[( he economizer-evaporizer-superheater (EES) g the reheater)

_OPERABLg pd "D

l b. For OPERABLE helius circulators, the following safe O P E. R A B L E shutdown cooling drives and auxiliary equipment shall be OPERABLE:

fie w p a't hs

1) A safe shutdown cooling drive with the capability of providing the equivalent of 8000 rps circulator speed at atmospheric pressure to two circulators f simultaneously, et t \ A y 7 # [n 2) Jw6 safe shutdown cooling drivef with the capability of providing 3% ra.ted helium flow at operating 3QC g $, M Ug pressure with firewater supply, including two d\

Os -

OPERABLE emergency water booster pumps and OeERABLE fiow Paths.

two

' 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

~

Ar.endment No.

Page 3/4.$- l DRf:-

[] b) At least FEBe 10% of prima ry coolant pressure nT v e boundary bolting and other structural bolting which W has been rsmoved for the inspection above M A and shall which is exposed to the primary coolant be nondestructively tested r g for identification of inherent or developed defects.

"Q

$L c) Reports v

or Within 90 days of examination completion, a N Sh Special Report shall be submitted to the NRC in !

(

accordance with $ specification 6.9.2. This

& I report shall include the results of the helium i g c. circulator examinations, nd. "

b.

The steam generators shall be demonstrated OPERABLE:

v

-t y 1. At .least once per 18 months be n, through the emergency feedvater/ verifying proper flow 8

header and emergency y condensate header to the steam generator sections. j

! t.

gj(.

l J ' # At*1 east once per 5 years by volumetrically examining

YD the accessible welds f portions of the following binetallic U < -~ 1 4 $ or indications of subsurface defects- t 9, ) 7).

b)l) The . main steam ring header collector to !

,r 3 , collector drain piping weld for one steam generator module in each loop, and f ..f n{ c. 8) ~ The same two steam generator modules shall be re-examined at each interval, v a. w

.c "

i F Q .f The initial examination shall be performed during SHUTOOWN cycle 5. or REFUELING prior to the beg. inning of fuel This initial exastnation shall also include the bimetallic welds described above for two additional steam generator modules in each loop. :

y, /. Tube Leak Examination .

Each time a steam generator tube plugged due to a leak, specimens from the accessible 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 preytous tube leaks. all A t ta a ,-t o n < e y e t- te m..rus by ver-11y : 3 tu D O PE R A B I.1-I.T Y o f ea c h su ret h eale r- o utle t ( few p aTh b yv e r-i -f y ' n 3 .T h o r r h, v alve s i ., t h e. stw ischpenT 1;nss c. a w b e o p en e el u d Th a T T h e, e nT l-lo w p a~1 b c 2re h o'r o b a T r u e T eel

Amendment No.

Page 3/4.5-SAFE SHUT 00VN COOLING SYSTEMS

~

3/4.5.1 SAFE SHUTDOWN COOLING EOUIPMENT g4-FEB 2 ~-

LIMITING CON 0! TION FOR OPERATION 3.5.1.2 a. At least one primary coolant loop shall be OPERABLE, including at least:

1. One helium circulator OPERABLE, and

2. One steam generator section (e'ner the economizer-

, y j ,.

evaporator-superheater (EES) or reneater) OPERABLp one OPER Ar ts b. For at least one OPERABLE helius circulator, the following

, flow faTk emergency drives and auxiliary equipment shall be OPERA 8LE:

I .g safe shutdown cooling drive with the capability of providing 3% rated helium flow at operating pressure with firewater supply, including one OPERABLE emergency i water booster pump and one OPERABLE flow path, 2g. The

)

turbine water removal system, including one turbine water removal pump,

.FOR 2 /. The normal bearing water syste., inciudin, one source of

,% sig bearing (P2105 water g P2108), makeup and and one bearing water makeup pump ,

O Q. ' ' 4g. The bearin t * ~

g T-2115) g water accumulator (T-2112, Tt2113. T-2114, .

for the OPERABLE circulator (s).

APpt!CABILITY: STARTUP*, SHUTOOWN=, and REFUELING

=

Vhenever CALCULATED 760 degrees F. BULK CORE TEMPERATURE is less than or equal to

Amendment No.

Page 3/4.5-C.9 A.

l ACTION: a. With less than the FEB 2 A

with forced circulation maintained, be in atabove required OPERA wiry , to a.+ 'sieiE, and restore the required equipment least SHUT 00VN to OPERABLE

"'"^'# status prior to reaching a CALCULATED BULK CORE TEMPERATURE K of 760 degrees F, or suspend all operations involving CCRE ALTERATIONS,. c.ontrof roct,wovemenTS resulting in positive re' activity changes, or, movement of JPAA01 ATE,0 FUEL.

E g. / bv eq )

4 t' L eq pe t an ar op n t ta t

' ^ nd V' o t s o p t re h ' A B C

/,s/ //

u t ta ie s v v *s es[oMvheMf I IA ,

2.

'\ '

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

$ SURVEILLANCE REQUIREMENT $ '

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

b.

With less than the above required OPERABLE forced equipment and with no fo minutes and suspend all operations cortrol rod involvingcirculation b!

CORE ALTERATIONS, movement of IRRADIATED FUEL, and either: movements resulting in positive 1

1.

Restore forced circulation on at least one loop prior to reaching a and comply with ACTION a, orCALCUIATED BULK CORE TEMPERATUR l i

wyn pgc to p rm .. c -

Specification 3.0. rovides the methodology and necessary data to determine the appropriate time interval to reach a CA14U1ATED BUlX CORE TEMPERATURE of 760 degrees F.

l l

n.

Amendment No.

Page 3/4.5 Oepressurization '

DiW[

g g g .g.g In the unlikely event that all forced circulation is lost for 90 minutes, start of depressurization is initiated as a function of prior power levels, with 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months from 1005 RATED THERMAL. POWER being the most limiting. case. Opera tors will continue attempts to restore forced circulation cooling until 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after the loss of forced circulation. Multiple sources and flowpaths to estabitsh forced convection cooling using circulators makes required depressurization highly unlikely. Cooldown using forced circulation cooldown is preferred to a depressurized cooldown,with the PCRV liner cooling system. Depressurization of the PCRV under extended loss of forced circulation conditions is accomplished by venting the reactor helium through a train of the helium purification system and the reactor building vent . stack filters to atmosphere. Start of depressurization times from various reactor power conditions 'are delineated in Figures 3.5.1-1, 3.5.1-2, and 3.5.13 and are discussed in the FSAR Section 9.4.3.3 and Appendix 0.

Redundancy Criteria The use of 760 degrees F CALCULATED BULK' CORE TEMPERATURE as a division between the APPLICABILITY of Specification 3.5.1.1 verses 3:5.1.2 is~ explained as follows:

In the FSV HTGR, the limiting parameter of interest is a core inlet temperature greater than 760 degrees F. The CALCULATED BULK CORE TEMPERATURE is a conservative calculation of the maximum potential temperature in the core and surrounding components. The conservatism are such that if the CALCULATED.

BULK CORE TEMPERATURE is ilmited 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 8ULK 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 accident failure consideration, prevention and mitigation without single cooling requirements. on the basis of the limited core rt G W W P S C. T o Pro #IM Specification 3.0hprovides the methodology and necessary data to determine the appropriate time interval to reach a' CALCUIATED BUlX CORE TEMPERATURE of 760 degrees F. If the active core remains below' this temperature, which. corresponds to the design maximum core inlet 3 i

temperature as indicated above, then the design core . inlet temperature cannot be exceeded and there can be no damage to fuel or PCRV internal components regardless of the amount, including total absence, or reversal, of primary coolant helium flow.

l

._ j

, .. l Amendment No.

Page 3/4.5-b5/.ET A I

\

All forced circula tion may be IE0 2 O E36 ,

I interrupted for maintenance purposes provided that the time calculated for CALCULATED BULK CORE TEMPERATURE to reach 760 degrees F is not exceeded.

However, if forced circula tion is temporarily res tored, a recalculation shall be performed, based on present conditions, l to establish a new time period for CALCULATED BULK CORE l TEMPERATURE to reach 760 degrees F. j also be taken out of service for maintenanceRedundant systems may testing provided or surveillance i 1

The time to reach CALCULATED degrees F may be recalculated as often as required.

BULKequal COREto 760 TEMPERAT 1 l

i Steam Generators Datel t 2d Th std g er rehe psc n,th Rcrinu- ome .thefthe s cia ed EES gen cti/-ns an ec ve t 1 em ea w , tu ater ead wh are r r

T. p r e v , J c. l per i pec, at

..r be l

n. stem gm, b ve ify P ITY s te j l

fl fr ch of t for e on d r

\

ead st o h ch am g ra r y f m graph en < itich I Bimetallic Weld Examination '

The steam generator 2nd crossover tube bimetallic welds between Wyo4!9 Incoloy 800 examination. and 21/4 Cr-1 No materials are not accessible for !

The bimetallic welds between steam generator gg1 ring header collector, the main steam piping, and the 2 m at collector drain piping are accessible, y rit materials, and operate at conditions not involve the same !

significantly different collector drainfrom theweld piping crossover tube bimetallic welds. The is also geometrically t

the crossover tube weld. Although similar to l

expected defects to occur, this specification allows for detectionminimal which of degradation i might affect bimetallic welds made result from conditions that can uniquely between these Additional collector welds are inspected at thematerials. initial examination to establish should defects be a baseline which could be used, examinations subsequently be required.found in later inspections and add Whenever the CALCULATED BULK CORE TEMPERATURE both the reheater and EES sections of theexceeds steam 760 degrees F, generator must be OPERABLE. The steam generator reheater water from either the emergency condensate or EES sections can receive f feedwater header as required to be OPERABLE per thisheader or the emergency per Specification 3/4.5.3. System Specification and verifying flow from each of the aforementionedflow OPERABILITY is determined by 140 3.5.3.1) through each section of each steam generator. emergency headers (see CALCULATED or the BULK CORE TEMPERATURE is less than or equal to 760 dWhenever the egrees F i determined headers (see LCO by verifying 3.5.3.2) flow from either of the gency aforementione generator. through either section of either steam j

___A- * " " ^ ^

s . + t ', : ,

Page 3/4 5-26

DRAFT SAFE $$LTDOWN COOLING SYSTEMS 3/4.5.3 EMERGENCY CCNDENSATE AND EMERGENCY FEE 0 WATER H i EMERGENCY CONCENSATE AND EMERGENCY FEECWATER -%. WEACERS -

LIMITING CCNDI7*0N F:9 COERATICN 3.5.3.1 The header shall be CPERABLE. emergency concensate header and the emergen APPLICABILITY: POWER, LCV PCWER, STARTUP", and SHUT 00VN" ACTION:

With either the emergency condensate header or the emergency feedwater header inoperable, restore the inoperable header l to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or: j

1. When in POWER, LOW POWER, or STARTUP, be in at least SHUT 00WN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or

2. When in SHUT 00WN, suspend all !

control red movement resulting operations involving in positive reactivity changes, or movement of IRRADIATED FUEL.

l t

( ., i SURVEILLANCE REQUIREMENTS _

?.5.3.1 No accitional Surveillance Requirements are required other tnan those surveillance icentified in Specification '.0.0.F

+

4, L l . le b . I

'Wherever degrees F. CALCULATED BULK CORE TEMPERATURE is greater than 760 M

e

r . c t ,3.

Page 3/4 S-27 ~

DRAFT SAFE SHUT 00VN COOLING SYSTEy$

M3 0 m 3/4.5.3 EMERGENCY CCNCENSATE AND EMERGENCY FEE 0 WATER HEADERS EMERGENCY CONDENSATE AND EugAGENCY FEE 0 VATER HEADERS ..- - - SEL!IM

. l LIMITING CCNCITICN POR OPERAT'ON 3.5.3.2 Either the condensate header or the emergency feedwater header shall be OPERABLE.

APPLICABILITY: STARTUP", SHUT 00WN", and REFUELING -

ACTION: Vith ot, th er ncy f dwat and h der opera e, tgency 94 densa sta s estor a e st one a e tg 0 LE io to *im calc' ed fo the c fronk ecay he to each g to he t CULATED ' LK COR TEMP nA U 760 deg F an in spen all e ations ivi' CC t A ~~* TIONS o ent I ro vement es ting i p reac i EL/od ti ty chang s, or move ent or RRA0! ATE 0' i a SURVEILLANCE REQUIREMENTS N .[#

4.5.3.2 No additional than those surveillanceSurveillance Requirements are required other 2 * ? .1 : .T identifiec in Specification q . T ,l 1. b , l -

Whenever 760 degreesCALCULATED F. BULK CORE TEMPERATURE is less than or equal to P S C n e e4 5 T P r o v '* J ' "M ~

Specification 3.0h determine ides the methodology and necessary data to TEMPERATURE the of appropriate 760 degrees time F. interval to reach a CALCULATED BUIX CORE L.

p.- --

~ -

~ -.

{ Vith both the emergency feedwater and emergency condensate header inoperable, be in at least SHUTDOWN within 10 minutes and restore at least one header to OPERABLE status prior to reaching a CALCUIATED BULK CORE TEMPERATURE $of 760 degrees F and suspend all operations involving CORE ALTERATIONS, control rod movements resulting in positive reactivity changes, or movement of IRRADIATED FUEL.

-mer.:.e t 4o,

' Page 3/4 S-28 -

DRAFT BASIS FOR SPECIFICATION LCO 3.5.3 / SR a.5.3 NOV 3 0,2 The OPERABILITY of the emergency condensate header and the emergency feedwater header ensures recureant water supoly paths to the helium circulators and steam generators for SAFE SHUT 00WN COOLING of the plant. **

failure of the In the" e ven t"*Tf""*4 normal. feedwa ter I f rie, the availability of either the emergency feedwater or emerger.cy condensate lines provides acecuate shutcown capactifty. OPERABILITY of the aforementioned verifying flow headers is accompitshed daring SHUT 00WN by through each neader to tr.e steam generators i and helium circulators.

In the FSV HTGR, the limiting parameter of interest is a core inlet temperature greater than 760 degrees F. The l CALCULATED BULK CORE !

TEMPERATURE is a conservative calculation of the maximum potential temperature in the core i and surrounding components.

The conservatism are such that if the CALCULATED degrees BULK CORE TEMPERATURE is limited to 760 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 P$C criterion greater then whenever CALCULATED 760 degrees F. BULK CORE TEMPERATURE is However, when CALCULATED BULK h c.ed 5 CORE TEMPERATURE is equal to or less than 760 degrees it F, is acceptable to require only one OPERABLE system for To accident prevention and mitigation wit sut single failure

,' consideration, on the basis of tne ilsited core cooling c r o r s .: requirements.

" f e r eas !

All fo'rced circulation may be inter .:ted for maintenance SP ' c h ab" purposes provided that the time calculated for CALCULATED 7*O. g BULK CORE TEMPERATURE to reach .760 degree s F is not

. exceeded. However, if forced circulation is temporarily restored, a recalculation shall be performed, based on present conditions, to ' establish a new time period for CALCUuTE0 BULK CORE TEMPERATURE to -each 760 degrees F.

Redundant systems may also be taken out of service for maintenance or surve.illance testing peevided that forced circulation is maintained; and the time to reach CALCU MTED 1

BULK CORE TEMPERATURE equal (recalculatedasoftenasrequired. to 760 degrees F may be The emergency feedwater header is not normally placed in service until approximately 30% reactor oewe r, to prevent .

I unnecessary long-term wear of components associated with the emergency feecwater header.

Nevertheless it is still required to be OPERABLE during the aforementioned N00ES.

l l

e S

n... s . ..f * 's *, '

\

. page 3/ *

. NOV 3 01985 With degrees CALCULATED F: BULX CORE TEMPERATURE less than or equal 'to 76 I

a. WIth one of the. above required pumps and/or makeup cones

inoperacle, restore the inoperable sculpfe'nt to status witn OPERABLE or water suoply.la days or provice an alternate backup pumo The provisions of Specification 3.0.4 are not apolicao'e.

b. With t.ae inoperacle,SAFE SHUTCOWN COOLING water system otherwise establish a backup suppression purposes W h4a ?A h r s:2-system for fire SURVEILLAN RE00!REWENTS A

4.5.4.1 The SAFE SHUT 00VN COOLING water supply system shall be demonstrated OPERABLE: I a.

At least once per 7 days by verifying the contained water

- supply volume in each of the circulating water makeup pones, (

I b.

At least once per 31 days by starting the electric motor-criven fire pump and operating it for at least 15 minutes, j

i

m. f c. At least ence per 31 days by star *.ing each circulating

.' j water makeup pump that is not alrea:/ running.

I d.

j At'l' east once per 31 days by verify 9.g that each valve in the flow path, that is not locked, sealed, or otherwise ,

secured in place is in its correct position,

e. At least once per 12 months by performance of a system flush, f.

At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel, g.

At least once per 18 months by performing a system functional actuation of test which includes simulated automatic the system throughout its operating secuence. and:

1. Ve ri fying that the automatic valve in the flow path actuates to its correct position. ,,

2. Verifying that each pump (motor-drAirn and engine-crives) develops at least 1425 som at 119 psig, prior to reaching e CALCUIATED BUlX CORE TEMPERATURE $f 760 degrees F within a period of time not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . but P5c n e n ,( , T . P r. v t / c " Al ",

h Specification 3.0 provides the methodology and necessary data to determine the appropriate time interval to reach a CALCULATED BULK CORE TEMPERATURE of 760 degrees F.

2 1

Y'i T-l'i

, y un m I With two cut of three circulating water makeup pumps (noperable or with any one firewater circulating water makeup pond inoperable or a pump inoperable and the TEMPERATURE greater than 760 degrees F, a restoration 41.CULATED .

time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and pumps. is considered sufficient based on redundant flow patns However, with all circulating water makeup pumps, headers, or firewater pumps inoperable, a restoration time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months supply is specified, are lost. as all means of SAFE SHUT 00WN COOLING water j

j The surveillance !

that all ecuipment, water supplies, and flow paths willidentified!

OPERABLE as specified in order remain COOLING requirements specified above.to meet those ~ SAFE SHUT 00wH 3

/'The use of 760 degrees F CALCULATED BULK CORE TEMPERATURE as a hvision between the ACTIONS is explained as follows; fn the FSV HTGR, the Ilmiting parameter of interest is a core "inist

' temperature greater than 760 degrees F.

' CORE TEMPERATURE' is a The CALCULATED P5C m. mum potential temperature conservative calculation of the cc sonents. in the core and surrounding I hce4i l BULK CORE TEMPERATURE is Ifmited inlet temperature of to 760 degrees F the design 7' used for accident prevention and760 degrees F is not exceeded. Systems satisfy the mitigation are required to l

(. c, ro i s - }

CORE TEMPERATURE is greater than However, 760 degress' F. single

/

CALCULATED BULK when '

teferewee CORE TEMPERATURE is ecual to or less then 760 cegrees for accident F, it is acceptable to require only one OPERABLE system i 5f 8 MMin '*"5'' " prevention and mitigation without single failure I requirements.' " *"' D'd 5 *"'

li*it'd ' ' 'I'"5 All i forced circulation may be interrupted for maintenance j purposes TEMPERATURE provided that the time calculated for CALCULATED BULK to reach 760 degrees F, is not exceeded. However, if forced circulation is temporarily restored, a recalculation l snal) be performed, based on present conottions. to establish a

new 760 time period for CALCULATED BULK CORE TEMPERATURE degrees F. to reach Redundant systems may also be taken out of service for maintenance or survet tlance testing provided that forced circulation is maintained; and the time to reach

( recalculated as often as required. CALCULATED BULK be CORE T

Amendment No.

Page 3/4 6-DR;h Tf PCRV AND CONFINEMENT SYSTEMS FEB28p'cc ;

3/4.6.2 REACTOR PLANT COOLING WATER /PCRV LINER COOL SHUTOOWN I

I LIMITING CON 0!TIONS FOR OPERAT!ONS_

l 3.6.2.2 I ""

APPLICABILITY: STARTUP

$NUTDOWN[*andREFUELINf ACTION: it o RPCW C

' i P I L

s t 1

pr o T es or at en a in a oe 1"5'"

CO P E O e p i er D g n o '

se en t ity c ng/ f SURVEILLANCE REQUIREMENTS __

/

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

1 Whenever CALCULATED to 760 degrees F. BULK CORE TEMEPRATURE 1s less than or equal

[Th c e5 o on f y e ut h in in s .ay s

s a E a o E s i

" s n e a o n a. J c T* fr*v:dc. "n ?

k Specification 3.0 @ g PS determine provides the methodology and necessary data to the appropriate time interval to reach a CALCUIATED BUIX CORE TEMPERATURE of 760 degrees F.

t-U M ih FO '

OhiY

6

)

L c o 3. 4, 2, 2 j l

I n s e ,- T A i

I The Reactor Plant Cooling Water (RPCW)/PCRV Liner Cooling System (LCS) shall be OPERABLE with:

a. One RPCW/PCRV LCS loop OPERATING with at least one heat exchanger' i and one pump in each loop OPERATING, and

b. With firewater supply.available via one OPERABLE flow path, j I n s er T G j With no RPCW/PCRV LCS loop OPERATING, within /0 minutes, be in at least i j

SHUTDOWN and suspend all operations involving CORE ALTERATIONS, control i rod movements resulting in positive reactivity changes, or movement of IRRADIATED FUEL, and:

a. RestoreatleastonelooptoOPEyTINGstatuspriorto reaching a 1

CALCULATED BULK CORE TEMPERATURE of 760 degrees F, or

b. tion cooling prior to reaching a CALCULATED Restoreforcedcirculgof760degreesF.

BUlX CORE TEMPERATURE l

l l

Amendment No. ;

Page 3/4 6-DRMT The use of 760 degrees F CALCULATED BULK CORE fra a e "EMPEllAfu'M6s a l division between the APPLICA8ILITY of Specification 3.6.2.1 3.6.2.2 is explained as follows: and I In the FSV HTGR, f$g inlet temperature greater than 760the Ilmiting parameter of interest is a core degrees F. The CALCULATED a BULK CORE maximum TEMPERATURE potential temperature in the is core a conservative and calculation of the h e 645 components. surrounding The conservati sms are such that if the CALCULATED Te BULK CORE TEMPERATURE is Itatted to 760 degrees F, the design inlet temperature of 760 degrees F is not exceeded. Systems used d"U~ for accident prevention and mitigation tre required to satisfy the refe r4Hef single failure criterion whenever CALCULATED BULK CORE i TEMPERATURE is grea ter than 760 degrees F. However, wheri l CALCULATED BULK CORE TEMPERATURE !

gfCI4'. '*T.*,, degrees F, it is acceptable to require only oneisOPERABLE equal to system or less than 760 1 3*O'p* required for accident prevention and mitigation as acceptable without single core cooling failure consideration, on the basis of the limited requirements.

1 All forced circulation say be interrupted for maintenance l purposes CORE provided that the time calculated for CALCULATED BULK i However, TEMPERATURE to reach 760 degrees F is not exceeded. i if forced circulation is temporarily restored, a l recalcu14 tion can be performed as required based on present plant conditions, tp establish a new time period CORE TEMPERATURE to reach 760 degrees F. for CALCULATED BULK Redundant systems say also be testing takenthat provided outforced of service for maintenance circulation is maintained. The or surveillance ;

time !

to reach CALCULATED BULK CORE TEMPERATURE equal to F760 degrees may be recalculated as often as required.

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AmendmenS No.

Page 3/4.7-DRAFT PLANT AND SAFE SHUTDOWN COOLING SUPPORT IESSYSTEMS 2 8 633

_3/4.7.1 TURBINE CYCLE 5AFETY VALVES - SHUT 00wN LIMITING CON 0! TION FOR OPERATION i

3.7.1.6 The steam generator superheater or reheater safety valve s which protect the OPERATING section(s) of the steam genera (or) shall be OPERABLE with setpoints in accordance with Table t 4.7.1-1.

APPLICABILITY: $HUTDOWN and REFUELING ACTION: With less than the above required safety vgive(s) OPERABLE, restore the required safety valve (s) to OPERABLE status prior F or suspend all operationsto reaching a CALCULATED BUL involving CORE ALTERATIONS or con t changes. rol rod movements resulting in positive reactivity SURVEILLANCE REQUIREMENTS.

4.7.1.6 No additional surveillance per Specification 4.7.1.5. required beyond those identified PSC need c T = p rev I M e

NI Y

Specification 3.0@ provides the methodology and necessary data to determine the appropriate time interval to reach a CALCUIATED BUlX CORE TEMPERATURE of 760 degrees F.

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!NFO ONLY

. Amendment No. 1 Page 3/4.7- l

' DRAFT l FEB 2 8 $80 l BAS 15 FOR SPECIFICATIONS 3.7.1.6/SR 4.7.1.6 LC0 3.7.1.5/SR 4.7.1.5 AND t.C0 ;

j The economizer-evaporator-superheater (EES) section of eacn steam l i

generator loop is protected by three spring-loaded safety valves, each with one-third nominal relieving capacity of each loop. The reheater section of each steam generator loop is protected from 1 overpressure transients by a single safety valve. These stean generator 10.2.5.3. safety valves are described in the FSAR, section 3

The above valves are required to be tested in accordance with (ASME Section XI, maintenance.

IGV requirements) every 5 years or after To satisfy the testing criteria, the valves must be .

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

as totominimize Due the operation at power untti the valves are tasted.

infrequent required testing of these valves, the likelihood of an accident oqcurring without proper valve i is considered vary small and plant safety is not compromised. testing {

During all MODES, with one EES safety valve inoperable, plant operation is restricted to a condition for which the remaining safety valves have sufftetent relieving capability to prevent ove' pressurization of any steam generator section (i.e., one botier feed pump per operating loop). Conversely, wit.h any j

reheater safety valve inoperable, plant operation is restricted

. to a more restrictive Mode.

A 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action time for repair or SHUTDOWN due to inoperable safety valves ensures that these valves are returned to service in a relatively short period of time, during which an j

overpressure hours does nottransient is unlikely. Operatton at pcwer for 72 for any extended period. result in a significant loss of safety function The setpoints for the safety valves identified in Table 4.7.1-1 are those valves identified in the FSAR with tolerances applied such that the Technical Specifications incorporate an upper bound setpoint. This is consistent with not incorporating normal operating limits in these Specifications.

, zo pTovide. t 1T * * * * \ ' 5'"

c Atc vt AT E D B U L i<. c_ o R E T E NI P E R -

Os.<nS A T V R E. BwM ~C o C ro f f -feherenet g pec (4 ; c a r l= " 3 O' '

~enemer.g No,

, Page 3/4 7-21 PLANT AND SAFE SHUTOCWN COOLING SUPPORT SYSTEMS 3/4.7.4 SERVICE WATER SYSTEM-SHUTOCWN NOV 3 01985 LIMITING COND_I, TION FOR OPERATICN - l l

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3. 7.4 ? The service water system shall be OPERABLE with:

a. Cae. service OPERABLE, water pump (P-4201, P-4202, or P-42025)

b. An of OPERABLE flow path to SAFE SHUTOOWN COOLING users service water (emergency diesel instrument coolers, air compressor and after coolers, and the l reactor plant cooling water /PCRV liner cooling heat i exchangers; E-4601, E-4602 E-4603, and E-4604), and 1

c. .

An CPERABLE flow path from the circulating water makeup system to the service water pump pit.

APPLICABILITY: STARTUP*, SHUT 00WN", and REFUELING ACTION:* With any of the requirements identified in 3.7.4.2 a, b, (c_ , ' or e above inoperable, restore the ecuipment to OPERABLE j status or r to reaching a CALCULATED BULK CORE TEMPERATUR involving CORE of 760 degrees F and suspend all operations ALTERATIONS o r. control rod movements 1

resulting in positive reactivity changes.

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'Vith theF.CALCULATED BULK CORE TEMPERATURE 1ess than or equal to 760 degrees t

f P S C- wecd T, p r o v ,* el e " A/

h Specification 3.0 n . .s.ine provides the methodology and necessary data to TEMPERATURE of 760 degrees F.the appropriate time interval to reach a C 1

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MMI h 4 7-h4 m 3 o w.o During POWER, 1 BULK CORE TEMPERATURE greater thanLOW POWER, STARTUP and SHUT 760 degrees F, with only one service water pump OPERABLE, .4 restoratt'on time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is provided to restore another pump to OPERABLE status. During this 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , service water needs can be met by the redundant pump.

With an inoperable flow path to the emer instrument air compressors and after coolers, gency diesel coolers or l cooling is required, initiation of backuo because the affected comconent may be automatically initiated tp perform 5AFE SHUTOOWN COOLING.

tnoperable flow path With an I

cooling heat exchangers, backup cooling capability must beto the reactor pj verified, so that it can be manually initiated if recured. Actual inittation j of backup cooling to the PCRV LCS heat exchangers is undesirable !

except in an actual emergency to minimize poss1bility of tube {

fouling. Seventy-two hours is provided to restore the flow path to 4 OPERABLE status, thts is consistent provided for in the system OPERABILITY with the restoration times components. requirements for the above If the flow path cannot be restored within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , j backup cooling will be initiated >

verified within (firewater) or the capability Both firewater pumps must be OPERABLE.1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to restore the equipment t for The surveillance Requirements 4 firewater pump OPERABILITY are given in Specification 3.5.4 If neither of these conditions can be met,~the plant must be in SHUTOOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If only one firewater pump is OPERABLE, a second i

pump must be restored to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or the plant must t

[ - be in SHUT 00WN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restoration time is required because service water is un ava.il abl e , and the firewater supply has no reduncant capability. If- no firewater pumps are ;

i OPERABLE, the plant must be in SHUT 00WN witnin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months because !

service water and backup cooling (firewater) are both unavailable.

With the flow path from the circulating water makeup system to the service water pump pit inoperable, a restoration time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is

provided to restore the flow path to OPERABLE status. This ACTION time is adequate, since backup cooling (firewater) can be and

(

considering the makeup requirements of the service water system. initiated, wcels During STARTUP, SHUT 00VN, and REFUELING, with the CALCULATED BULK 7' CORE pump, TEMPERATURE less than or ecual to 760 degroes F, a service water cro f f -

a flow path from the circulating water makeuD system toflow p the service W 9.g water pump pit are recuired to be OPERABLE prior to the time calculated F. to reach a CALCULATED BULK CORE TEMPERATURE of 760 deg b f , g ,., g ;,, This ACTION ensures that the plant will remain in a stable y ,9, Q , condition testing, or when service outages.

unanticipated water is unavailable due to maintenance,

~

No specific surveillance are required as the service water system is normally flow pathoperating, including the required OPERABLE flow paths. The by the surveillance testing of the diesel generators.to the stancey die

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(; ENCLOSURE 5 Categorization of the Auxiliary Electrical Power Systems, 3/4.8, connents as transmitted to PSC by the NRC Letter of May 6,1987, Enclosures 1 and 2. Categorization symbols are the same as in previous I

meetings and correspondence (for example, the C. Hinson memorandum of December 15, 1986). ,

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Conrnent No. _

Categorization _

ENCLOSURE 1 1 0 2 0 3 0 4 D 5 D 6 0 7 0 '

ENCLOSURE 2 l 1

1 B 2 0 3 0 4a B l 1

4b 8 4c D 4d B i 4e B j 5 0 6 0 7 D*

8 D 9 0 10 0 11 0 12 0 13 0 14a F 14b,c D .

l 15 0 16 D for reactivity

, change poriton 17 0 18 D 19 0 20 0 21 0 22 0 23 0 24 0 25 0 26 0 21 0 28a D*

28b B 28c D*

2 1

Conraent No. Categorization ENCLOSURE 2 (Continued) 28d B 28e B 28f A8 0A# for Licensee action to address 5 days vs 2 hrs, 289 0 29h 8

29) A#

29j B l

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v t e Letter Sequence RAI
TAC:56565, (Approved, Closed)
Initiation Acceptance... Administration Meeting, Meeting, Meeting, Meeting Questions 2 July 1987 Answers Results Approval, Approval, Approval, Approval Other: ML20205F038, ML20205F047, ML20206P200, ML20213D665, ML20236G302
MONTHYEAR1987-07-02 [Table View]

ML20235F336

Text

Dear Mr. Williams:

SUBJECT:

References:

Enclosures:

Dear Mr. Williams:

SUBJECT:

References:

Enclosures:

Dear Mr. Williams:

SUBJECT:

References:

Enclosures:

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