ML20203E149

Forwards Rev 4 to Updated Fsar.Rev Includes Effects of Changes Made to Facility or Procedures,Util Safety Evaluations & Analyses of New Safety Issues.Description of Changes or Additions Also Encl

ML20203E149
Person / Time
Site:	Fort Saint Vrain
Issue date:	07/22/1986
From:	Robert Williams PUBLIC SERVICE CO. OF COLORADO
To:	Berkow H Office of Nuclear Reactor Regulation
Shared Package
ML20203E154	List: ML20203E149
References
P-86466, TAC-56743, TAC-59314, NUDOCS 8607240070
Download: ML20203E149 (19)

Text

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2420 W. 26th Avenue, Suite 1000 Denver, Colorado 80211 July 22, 1986 Fort St. Vrain Unit No. 1 P-86466 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20555 Attn: Mr. H.N. Berkow, Director Standardization and Special Projects Directorate Docket No. 50-267

SUBJECT:

Fort St. Vrain Updated FSAR, Revision 4

REFERENCE:

1) PSC Letter Brey to Gagliardo dated July 22, 1986 (P-86454)

Dear Mr. Berkow:

Enclosed is one (1) signed original and twelve (12) additional copies of Revision 4 to the Updated Final Safety Analysis Report (Updated FSAR) for Fort St. Vrain (FSV) which has been prepared and is being submitted in accordance with 10CFR50, Section 50.71(e).

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

Brief descriptions of the principal changes or additions to affected sections and appendices of the Updated FSAR, Revision 4, are included in Attachment A. 10CFR50.71(e)(2)(ii) requires that this submittal identify changes made under the provisions of 10CFR50.59 but not previously submitted to the Commission. Most changes reflected in Revision 4 of the Updated FSAR are identified in the current 10CFR50.59 annual report (Reference 1). Other changes made in the FSAR, not requiring prior Commission approval pursuant to 10CFR50.59(a), and not yet reported to the Commission, are listed in Attachment B of this letter. These changes will be more fully described along with the results of their safety evaluations, as appropriate, in PSC's next annual report of 10CFR50.59 changes.

Attachment C contains a summary of PSC's resolution of comments q6g ,

I provided by the NRC on Revision 3 of the Updated FSAR.

8607240070 860722 PDR ADOCK 05000267 K PDR L

e P-86466 Pagg 2 July 22, 1986 If you have any questions or comments, please contact Mr. M. H.

Holmes at (303) 480-6960.

Very truly yours, ,

R. O. Williams, Jr.

Vice President, Nuclear Operations R0W/JW:jmt Attachments:

A - Description of Principal Changes Incorporated in Revision 4 of the Updated FSAR 8 - List of 10CFR50.59 Changes Not Previously Reported to the Commission C - Resolution of NRC Comments Concerning Revision 3 of the Updated FSAR

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r-UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter Public Service Company of Colorado ) Docket No. 50-267 Fort St. Vrain Unit No. 1 )

AFFIDAVIT R. O. Williams, Jr., being duly sworn, deposes and says that he is Vice President of Public Service Company of Colorado; that he is duly authorized to sign and file with the Nuclear Regulatory Commission Revision 4 of the Updated FSAR; that he is familiar with the the content thereof, and that the matters set forth therein are true and correct to the best of his knowledge, information, and belief.

4 E. O. Williams, Jr.

Vice President, Nuclear Operations STATE OF bm de )

)

COUNTY OF gUmw )

Subscribed and sworn to before me, a Notary Public on this

1. 1M/ day of @u /m , 1986.

u' J 0?n?utJ $ $ //% AN?ird I

"*lF/N"$1Nbmat/Nbiuu' t111tw, Ofo 9022/

My commission expires h u /rfc>l M , 198f J

Attachment A to P-86466 Page 1 DESCRIPTION OF PRINCIPAL CHANGES INCORPORATED IN REVISION 4 0F THE UPDATED FSAR Section 1 Some of the information associated with the former FSV environmental qualification program was removed. Statements were added regarding the extensive program underway to ensure that safe shutdown electrical equipment is in compliance with 10CFR50.49 " Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants." When this program has been completed and accepted by the NRC, the FSAR will be updated to reflect it. Table 1.4-2 was corrected to refer to either battery "1A" or "1B" instead of two out of three station batteries. The purification cooling water system was added to Item 10 of Table 1.4-1. The control room doors were revised on Figures 1.2-17 and 1.2-18 to reflect changes needed for fire protection requirements. The description of the prestressing of the PCRV was revised to reflect recent research and testing.

Section 3 Section 3 was revised to include changes made during the Control Rod Drive (CRD) Refurbishment of the CRDs, including: the change of control rod cables to a new material; addition of helium purge flow recorders, and the addition of remote temperature devices (RTDs) and temperature recorders; addition of knockout pots on steam generator and helium circulator interspaces. Those changes improve reliability and monitoring for the CRDs and help to prevent moisture ingress to CRDs and the reserve shutdown system.

Revisions were also made to better support the Technical Specifications. These included an improved discussion of chemical impurity limits of the primary helium and an improved discussion of core graphite oxidation rates at low power.

Discussions of the basis for the core safety limit (power-to-flow ratio), the reactivity worth of reserve shutdown (RSD) units with the highest worth unit inoperable, and a clarification of control rod worth versus withdrawal distance were also added. Additional references were added that better explained the ability of fuel particles to withstand temperatures well in excess of 2900 degrees Fahrenheit. Technical Specifications surveillance testing criteria were added for CRD inoperability, control rod drop test, CRD motor temperatures, RSD hoppers, RSD material, and CRD orifice assembly preventive maintenance. The control rod scram time was revised to be consistent with that used in the accident analyses of 152 seconds.

The discussion of the modification to the orifice valve control circuitry was revised and an additional analysis of a stuck orifice valve condition was added.

Attachment A to P-86466 Page 2 Section 4 An additional discussion was added to clarify the purpose and operation of the helium circulator nitrogen pressurization system. The remote manual control of the circulator brake and seal for fire protection was added. The reason for moisture limitations in the primary coolant was more completely explained.

The discussion of main steam outlet temperature resulting from postulated rod withdrawal accident was revised due to the removal of the feedwater latch circuitry. There was a revision to the discussion on backup bearing water due to a change in location of the pressure sensing point to provide for more stable operation.

The removal of unnecessary bearing water pump discharge check valves was also documented.

Section 5 Revisions included the incorporation of new developments in PCRV tendon corrosion. Discussions of minor modifications of tendon end caps, drain valves and tendon 0-rings were also included.

The Interim Tendon Surveillance Program discussion has been enhanced. Actions to be taken if a tencon load cell alarms and tendon failure criteria were added.

The discussion of PCRV helium leak rates was expanded and the description of maximum helium leak rates was improved to resolve an apparent inconsistency between the FSAR and the Technical Specifications. Volumetric leak rates were clarified as being

" standard" whenever applicable.

The installation of knockout pots on steam generator and helium circulator interspaces was also noted.

Section 6 Testing requirements of reactor building exhaust filters were clarified.

Section 7 Included the effect of modifications to dewpoint moisture monitor detectors to allow for better maintenance and remove the scattered light feature which has never been operational, as well as the re-routing of moisture monitor tubing within PCRV penetrations to provide easier access. Revised the discussion of the orifice valve control circuitry to incorporate the modification which provides for a continuous actuation mode.

The explanation of detector decalibration was expanded. Figure 7.1-11 was revised to reflect circulator speed /feedwater flow program changes. The definition of the Reactor Protection System and its relationship to the Plant Protective System was

. _ _ _ ~.

Attachment A to P-86466 Page 3 clarified. A description of the power-to-flow ratio measurement was added.

Section 8 Cables for the diesel engine and the water jacket heaters were revised to reflect the upgrading of the cable material. Building 10 fire detection / suppression and emergency lighting systems were revised to show the new connection to a more reliable electrical source. Figures 8.2-14 and 8.2-20 were updated to show the relocation of the power feed to the computer data logger static inverter.

The independence of the emergency diesel generators and application of the single failure criteria was clarified. The fact that essential loads are de-energized before being energized again by the emergency diesels was stated. There was a revision to indicate a change to valving in cooling water for the High Temperature Filter Adsorber (HTFA) for cooldown using the Alternate Cooling Method (ACM) system. A correction was made to state that the diesel generator overload relays and antimotoring relays are blocked in the auto-start mode. Also added was the battery nameplate and four hour profile battery rating, as well as clarification that the batteries are capable of supplying shutdown DC loads for four hours. It was noted that Flamemastic 77, in addition to Flamemastic 71A, is acceptable as a fire retardant.

Section 9 Revised to include the addition of the outside liquid nitrogen section, consisting of two 13,000 gallon nitrogen storage tanks, two liquid nitrogen pumps, and associated piping. The addition of piping to permit the use of the recondenser chiller's colder wcter for cooling the helium purification coolers was reflected.

Six flow switches and their associated alarms were added to the fixed water spray Fire Stations in the areas of the "G" and "J" walls. A revision was made to note the addition of permanent piping and valves in place of spoolpieces for cooling the HTFAs.

New waste lines to allow system 61 to receive, store, and process liquid waste from systems 23 and 62 were reflected. Addition of the auxiliary helium tube trailer was incorporated.

Discussions on the flow rate of air required for cooling of spent fuel storage wells were clarified. Results of the heavy loads analysis Phase I and Phase II were included. The limiting conditions for operation of the liner cooling system was added.

It was noted that Flamemastic 77, in addition to Flamemastic 71A, is acceptable as a fire retardant. A discussion of the "G" and "J" wall congested cable area closed head water spray system was addad. A summary of the evaluation of compliance with 10CFR50, Appendix R, and an updated comparison of Appendix A to BPT-9.5-1, which was previously submitted to the NRC, was added.

Attachment A to P-86466 Page 4 Section 10-Revised to reflect the safe shutdown cooling demands on the capacity of the storage ponds. Deleted specific reference to hydraulic or mechanical shock suppressors for interchangeability.

Corrected and clarified pipe specifications for main steam, reheat steam, feedwater, and condensate lines. Clarified the use of the Emergency Water Booster Pumps for emergency cooling.

Section 11 Revised to add control changes which cause the shutdown of the liquid waste transfer pumps and the reactor building sump pumps upon detection of high gama activity. Clarified to indicate that both helium purification coolers and regeneration knockout pots drain to the decontaminant solution tank or the liquid waste receivers.

Section 12 Updated the organization description and added the acceptance of the temporary use of R0s in place of SR0s for licensed operator requalification for 1985-1986. Section 12.3.6. was added to provide a description of " Post Trip Reviews".

Section 13 The discussion of the rise to power testing (70-100%) and region peaking factor data update requirements following each refueling was revised.

Section 14 Revised to include re-analysis of reactor building temperature and pressure following MCA and pressures following DBA-2.

Clarification was provided concerning the use of two feedwater-supplied, Pelton wheel-driven circulators, conservatively assumed to be in the same loop for the DBA-2 analysis. Analyses of temperature transients for 30 and 60 minute delay for start of cooldown following a Design Basis Depressurization Accident was inserted. A description of the probability of DBA-2, with the addition of radiation levels following DBA-2, was added. Revised the SCRAM reactivity for a rod withdrawal accident. Revised the CRD scram time to 152 seconds to be consistent with the accident analyses. Included description of relocation of circulator pressure differential elements to the exterior of PCRV penetrations. A discussion was added on operator action required when reheater is flooded to prevent exceeding PCRV rupture disc setpoint. The malfunctions involving handling of heavy loads were added. Clarification was made of the analytical assumptions of air flow for emergency cooling of the fuel storage vaults.

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Attachment A to P-86466 Page 5 Appendix A A new section was added to discuss the results of investigations of chloride contamination of the primary coolant circuit. The results of recent surveillances of fuel test elements and fuel surveillance elements were also added. A new paragraph added to describe the local and remote manual actuation of the helium circulator brake and seal system.

Appendix B Revisions to update the organization were included.

Appendix C Clarified to indicate actions not controlled from the control room. It was verified that the reserve shutdown system is not automatic. The capability of interruption of coolant flow for at least 1 1/2 hours (instead of thirty minutes) was discussed. It was also clarified to indicate that the reactor can withstand an interruption of forced circulation cooling for at least 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> before unacceptable damage to the steam generator inlet ducts would occur upon circulator restart. The "G" and "J" wall congested cable area fire protection systems modification to a closed head water spray systems was described.

Appendix D Description was added of the newly installed pipe and valves (in place of the temporary spoolpiece) for ACM using the HTFAs. The discussions concerning the PCRV helium leak rates was corrected.

Core temperatures and heatup connected with DBA-1 were clarified.

Appendix H The list of class I piping in Appendix H was modified to include new piping installed for nitrogen system changes.

Attachment B to P-86466 Page 1 i

LIST OF 10CFR50.59 CHANGES NOT PREVIOUSLY REPORTED TO THE COMMISSION As required by the 10CFR50.71(e)(2)(ii), following is a list of the changes made under the provisions of 10CFR50.59 which have not been submitted to the Commission. These changes will be more fully described, along with the results of their safety evaluations , as appropriate, in PSC's next annual report of 10CFR50.59 changes.

FSAR SECTION DESCRIPTION OF CHANGE

1. Revised the control room doors on Figures 1.2-17 and 1.2-

18. Changes were made to heavy load handling equipment.

3. Added knockout pots, moisture elements, pressure transmitters, independent backup source of dry helium, and associated instrumentation for CRD purge line and reserve shutdown system. Changed various CRD0A tolerances.

4.2 Removed the feedwater latch reset circuitry.

7 Changed Figure 7.1-11 to reflect circulator speed /feedwater flow program changes. Modified the orifice valve control circuitry discussion to provide a continuous actuation mode.

8. Relocated the power feed to the computer data logger static inverter N-9255. Changed the electrical connection for Building 10 emergency lighting and fire detection / suppression system.

9. Permanent piping and valves have been installed so that a temporary spoolpiece does not have to be installed for emergency cooling of the HTFAs.

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' B. The component data bases have additional identifiers to indicate those Security and Fire Protection components that are not safety related, but to which portions of the quality assurance program are applied.

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Attachment C to P-86466 Page 1 RESOLUTION OF NRC COMMENTS CONCERNING REVISION 3 0F THE UPDATED FSAR

1. NRC Connent: Letter G-86008 (Berkow to Walker dated December 27, 1985) concerning PSC's Technical Specification Upgrade submittal (P-85363) lists as Item 5 "The FSAR should be updated to clarify the licensees position on circulator interlocks whose failure could prevent any source of motive power from being supplied to circulator drives."

PSC Response: Sections 4.2.2.1. and 4.2.2.3.5. have been updated to include a description of the interlocks on the circulator inlet and outlet valves.

In Attachment 1 to letter G-85433 (Butcher to Lee dated October 21, 1985) concerning the Technical Specification Upgrede, five discrepancies were identified.

2. NRC Comment: Section 5.12.2., Rev. 2 - The maximum helium leakage of 1 lb./hr. = 100%/yr. does not appear to be consistent with existing LC0 4.2.9. which allows a secondary seal leakage of 400 lb./ day.

The staff recommends that the FSAR be revised to resolve this inconsistency.

PSC Response: Section 5.12.2. has been revised to clarify the helium leakage rates.

3. NRC Comment: Section 6.2.3.2.3., Rev. 3 - This section implies that the Reactor Building exhaust filters comply with R.G. 1.52, Revision 2. The staff recommends that the FSAR be revised to clarify the exception taken to R.G. 1.52, Revision 2, concerning the frequency (i.e.,

4400 hrs. versus 720 hrs.) of laboratory

! analysis as outlined in the existing LC0 SR l 5.5.3, " Reactor Building Exhaust Filters,

( Survi.illance."

PSC Response: Section 6.2.3.2.3. has been revised to clarify the testing of the charcoal filter material.

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Attachment C to P-86466 Page 2

4. NRC Comn.ent: Section 7.3.1.2.1., page 7.3-9, Rev. 3 - This section refers to a Figure 7.3-24 which apparently was not incorporated into the FSAR.

FSAR, List of Figures, Revision 3, does not list a Figure 7.3-24. The staff recommends that the FSAR be revised to correct this inconsistency.

PSC Response: The reference should have been to Figure 7.3-20, instead of Figure 7.3-24. This reference has been corrected in the rewrite of Section 7.3.1.2.1.

5. NRC Comment: Section 7.3.5.3., Rev. 2 - This section and the basis for the proposed LC0 3.3.2.2. refer to a FSAR Table 7.3-2, which was apparently not incorporated into the FSAR. FSAR, List of Tables, Revision 3, refers to a Table 7.3-2,

" Process and Area Radiation Monitoring Systems

... (Rev. 1)." The staff recommend that this table be incorporated into the FSAR.

PSC Response: Table 7.3-2, " Process and Area Radiation Monitoring Systems ... (Rev. 1)" does exist in the Updated FSAR. No change is required.

6. NRC Comment: Section 9.7, Rev. 2 - Figure 9.7-1 shows only one liner cooling system pump per loop whereas, Figure 9.7-2 shows the two liner cooling system pumps per loop. The staff recommends that the FSAR be revised to correct this inconsistency.

PSC Response: Figure 9.7-1 has been revised to clarify that there are two liner cooling system pumps per loop.

An Informal Transmittal from the NRC on May 19, 1986, contained many comments on the FSAR, Revision 3.

7. NRC Comment: RT 93250-14 was added to Figure 1.2-5,

" Equipment Identification List," but was not identified on the reactor plant arrangement plan as being located next to N-9225 (CRD MCC#1).

PSC Response: Figure 1.2-5, " Reactor Plant Arrangements, Plan Above Elevation 4881 ft.", has been revised.

8. NRC Comment: The change to Section 3.6.6.1. inadvertantly deleted the discussion as to where the second ICRD was installed. Also, the abbreviation "ICRD" was defined but not used in the sentence, which stated in part, " Fluctuation

Attachment C to P-86466 Page 3 testing resumed ... Instrumented Control Rod Drives ...".

PSC Response: Section 3.6.6.1. has been revised to identify Region 35 as the location of the second "ICRD" and the abbreviation ICRD was utilized in the section.

9. NRC Comment: Section 3.8.4.2. and D.3.3.2. states that approximately 1/2 of the RSS absorber material was released instead of the approximately 1/3 previcusly reported to the NRC.

PSC Response: The correct figure is 1/2 of the RSS absorber material was released. The FSAR reference has been updated to include the final LER submittal.

10. NRC Comment: Section 3.8.4.3.1. does not list all the CRD0A modifications committed to by the licensee (e.g., clevis pin material replacement, limit switch cam modification, etc., identified in the licensee's letters dated January 31, 1985 (P-85040), March 1, 1985 (P-85061), and June 25, 1985 (P-85224)).

, PSC Response: Section 3.8.4.3.1. has been revised. This

. section will be further revised for clarity of specific components in Revision 5.

11. NRC Coment: Section 4.2.2.3.7., page 4.2-25 states that backup bearing water is obtained from steam turbine drive feedpumps only. This section failed to identify that backup bearing water is also supplied by the electric motor driven feedpump via the emergency feedwater header.

PSC Response: The condition addressed by Section 4.2.2.3.7.,

page 4.2-25, addresses a power failure situation in which the electric motor driven feedpump is not available.

> 12. NRC Comment: The word " surveillance" is misspelled at the top of page 5.6-8.

PSC Response: The spelling has been corrected.

13. NRC Coment: Figure 6.8-2 changes appear to be in error.

Section 3.8.3. implies that the low set rupture disc is in line with the low set safety valve and that the high set rupture disc is in line with the high set safety valve. However, Figure 6.8-2, Revision 3, has now reversed the

Attachment C s to P-86466 Page 4 rupture disc set points such that the low disc is in line with the high safety- and the high disc is in line with the low safety.

PSC Response: Figure 6.8-2 has been revised and is now

, consistent with Section 6.8.3. and drawing PI-l 11-5.

14. NRC Comment: Section 7.4.1. refers to a Figure 7.4-2 which does not exist. Apparently, this reference should be the newly added Figure 7.4-1.

PSC Response: The reference has been changed to Figure 7.4-1.

15. NRC Coment: The next to the last sentence of the first paragraph in Section 9.12.5.3. is improperly worded. A more appropriate choice of wording would be, "Since Section III.L contains the acceptance criteria for a portion of Section i III.G for Light Water Reactors, it was necessary to develop acceptance criteria for i Section III.G applicable to FSV." This would delete the inference that PSC developed their own acceptance criteria.

PSC Response: The language of Section 9.12.5.3. has been i revised to incorporate the suggested language.

16. NRC Coment: Section 9.12.5.3., page 9.12-13, third paragraph, contains the misspelled word "non-congested".

PSC Response: The spelling has been corrected.

17. NRC Coment:

The 2109licensee's and P-2110 attempt (booster to pumps clarify)theinuse Section of P-

. 10.3.9., page 10.3-8, was unsuccessful. As currently worded, it still is not clear that to i obtain one circulator's design speed for DBA-2, one fire water pump and one booster pump are required.

i PSC Response: Section 10.3.9. has been further revised to make clear that one emergency water booster pump is used in conjunction with firewater pressure. Use of firewater and emergency water booster pumps is not applicable to DBA-2.

t

18. NRC Comment: The second sentence of the first paragraph under Section 12.1.3.1. is redundant.

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Attachment C to P-86466 Page 5 PSC Response: The sentence has been deleted.

19. NRC Comment: As worded, Section 12.1.4.15 implies that there are more than one reactor engineer. The licensee should make the title singular and revise "One Reactor Engineer..." to read "The Reactor Engineer..."

PSC Response: The description of Reactor Engineer in Section 12 has been deleted since the Supervisor, Technical Services is performing the function of Reactor Engineer. Table 12.1-1 also reflects this fact.

20. NRC Comment: It was not possible to delete Table B.5-3, Revision 1, from the FSAR, since Table B.5-4 was on the back page and Table B.5-4 was to be maintained in Revision 3.

PSC Response: Table B.5-3 is on two sheets back-to-back as is Table B.5-4. No correction seems to be necessary.

21. NRC Comment: The change made to page 0.2-7 contains an editorial error (e.g., " lines barrel" should be

" liner barrel").

PSC Response: The spelling has been corrected.

22. NRC Comment: The subscript for Cp was inadvertantly deleted from the equation terms in Section 14.12.8.2.,

Revision 3.

PSC Response: The subscript has been corrected.

23. NRC Comment: Section 3.6.3.1., Page 3.6-4, Revision 3, revised the Cp equation to correct a problem with the third term, however, it would appear that a plus sign is still required after the 4.88E-06T squared term to correct the equation.

PSC Response: The equation has been corrected by inserting a minus sign.

24. NRC Comment: Item 14 of Table 1.4-2, Revision 3, list two of the three station batteries as required for a safe shutdown. This contradict 3 Section 8.2.3.4., Revision 3, which states that either battery 1A or 1B would be adequate to supply shutdown DC loads for not less than one hour following a loss of all AC power. Battery IC only provides a source of power to the 120 VAC non-interruptible instrument power Bus IC.

Attachment C to P-86466 Page 6 Section 8.2.5.2., Revision 3, indicates that the loss of a DC bus (e.g., single failure) does not result in an unsafe condition, since two separate, independent systems are provided (i.e., batteries 1A or 18).

PSC Response: Table 1.4-2, Revision 3 has been revised to read battery "1A" or "1B".

25. NRC Comment: Section 3.2.3.3., Revision 3, states that temperatures well in excess of 2900 degrees F, instead of the previous "up to 2900 degrees F",

may be withstood for short periods without rapid deterioration of the fuel particle fission product barrier. This revision was not justified by supporting information.

PSC Response: A reference has been added to include the safety evaluation indicating that temperatures well in excess of 2900 degrees F may be withstood for short periods of time.

26. NRC Comment: Table 3.2-2, Revision 3, revised the maximum rod insertion time for scram from 152 seconds to 160 seconds without referencing supporting justification.

PSC Response: The changes made to change the rod insertion time to 152 seconds were made to reflect the time used in the accident analyses.

27. NRC Comment: Section 3.4.1.2., Revision 3, added two paragraphs that discuss replacement of full length reflector blocks with either full length or half length reflector blocks without providing supporting justification.

PSC Response: The addition of this justification has been deferred.

28. NRC Comment: Section 3.8.4.3.3., Revision 3, discusses modifications to the CRD0A helium purge supply required for control rod drive and orifice assembly operability. These modifications have not been completed.

PSC Response: These modifications were installed by CN 1923 and CN 1967 which are complete.

29. NRC Comment: Section 4.2.4.1., Revision 3, removed the steam generator interspace rupture disc and relief valve discussion on Page 4.2-31 for no apparent l

Attachment C to P-86466 Page 7 reason.

PSC Response: This discussion was moved from Section 4.2.4.1.

to Section 5.8.2.5.4.

30. NRC Comment: Section 5.8.2.5.5., Page 5.8-12, Revision 3, changed the nonnal bearing water supply value from 100 gpm to 170 gpm without providing justification.

PSC Response: The change was made to agree with Section 4.2.2.3.2., page 4.2-10 and Section 4.2.2.3.3.,

page 4.2-12.

31. NRC Comment: The Section 6.2.2., Page 6.2-2, Revision 3 discussion (not revised) that equipment

" essential for safe operation of the plant, has been designed to operate satisfactorily during the brief period of high-temperature environment" appears to be inconsistent with present environmental qualification studies.

PSC Response: Discussion of present environmental qualification studies has been added to Section 1.4.5.4. and 1.4.6.

32. NRC Comment: Section 7.4.1., Revision 3, clarified the modes of operation for the control room ventilation system and should be referenced in the basis for the proposed draft upgrade LC0 3.7.9.

PSC Response: A future revision of the FSAR will be coordinated with the approval of the Upgraded Technical Specifications.

33. NRC Comment: Section 8.2.3.4., Page 8.2-13, Revision 3, indicates that the required service capacity for Batteries 1A and IB was changed to 1058 amp-hour (4-hour profile); in Revision 2 this value was revised to indicate a capacity of at least 832 amp-hour. While a number of pieces of correspondence have discussed battery capacity design and requirements there still appears to be some confusion on how much actual capacity is required.

PSC Response: The battery nameplate capacity and four hour profile loading are included in Revision 4.

34. NRC Comment: Figure 8.2-14 is not legible.

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Attachment C to P-86466 Page 8 PSC Response: Figure 8.2-14 legibility has been improved.

35. NRC Comment: Section 8.2.3.5., Page 8.2-14, Revision 3, changed the sequence of operations shown for a loss of all off-site power, with the turbine generator online, to state that no loads are connected to the diesel generators when the diesels automatically start. The previous revision indicated a manual, synchronized transfer of main turbine generator loads to the

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diesel generators. Revision 3 just states, "Following an essential load transfer, the plant may be shutdown." without explaining how the load transfer is accomplished.

PSC Response: The sequence of operations for loss of all off-site power is clarified.

36. NRC Coment: Section 9.12.6., Revision 3, does not list all the smoke detectors submitted in the proposed draft upgrade LC0 3.3.2.5.

PSC Response: A future revision of the FSAR will be coordinated with the approval of the Upgraded Technical Specifications.

37. NRC Coment: Proposed draft upgrade LC0 3.3.2.5. list the minimum number of operable smoke detectors in the auxiliary electric room air duct as one, whereas, FSAR, Section 9.12.6., Revision 3, list the number as zero.

PSC Response: A future revision of the FSAR will be coordinated with the approval of the Upgraded Technical Specifications.

38. NRC Coment: Section 14.1.1., Revision 3, deleted the IE Bulletin 80-11 references without justification.

PSC Response: Three references concerning IE Bulletin 80-11, Masorary Block Walls, have been added to Section 14.1.1.1.

39. NRC Coment: Sections 14.2.2.1. and 14.2.2.7., Revision 3, revised the lower rod withdrawal prohibit (RWP) setpoint form 25% to 30% without supporting justification.

Section 14.2.2.1., Revision 3, revised the higher RWP setpoint form 120% to 108% without supporting justification.

Attachment C to P-86466 Page 9 Section 14.2.2.1., Revision 3, revised the SCRAM setpoint from 140% to 116% without supporting justification.

Sections 14.2.2.1. and 14.2.2.7., Page 14.2-7 and 14.2-11, Revision 3, indicate a higher RWP

, of 120% rated power which does not agree with 3

the Section 14.2.2.1. maximum RWP trip setting of 108%.

PSC Response: Clarification has been added to the FSAR for these functions. Also, the safety evaluation submittal is added to the references for Section 14.2.

40. NRC Comment: Revision to Section 14.5.2.2. titled, " Core Support Post and Bottom Reflector Effects,"

appear to be in error. Further, no justification was referenced for this change.

For example, the fourth paragraph was changed to indicate that the steam attack worst case was computed for the core support post instead of the previous reference to core support blocks. However, the next sentence implies that the worst case computed values can be conservatively applied to the core support post by comparison to the attack on the core support blocks, which does not make sense if the worst case was computed for the support post to begin with. Also, the revision failed to address the source of the new values for the average fractional burnoff.

PSC Response: The word " post" has been changed to " block",

the section has been enhanced for clarification, and references have been added.

41. thru

44. NRC Coment: There were errors in the use of change bars for Revision 3.

l PSC Response: The change bars have been revised or superceded for Revision 4.

45. NRC Comment: Attachment A to P-85254 provided brief descriptions of the principal changes or additions to Revision 3. Attachment A implied that the actuation settings for the steam generator relief safety valves and rupture i

discs were clarified in Section 4. However, this change was incorporated in Section 5.8.2.5.4. instead of Section 4.

Attachment C to P-86466 Page 10 PSC Response: The comment is correct, the change was made in Section 5 instead of Section 4.

46. thru

52. NRC Comment: There were errors in the use of change bars for Revision 3.

PSC Response: The change bars have been revised or superceded for Revision 4.

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