ML20235J400
| ML20235J400 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 01/16/1987 |
| From: | Robert Williams PUBLIC SERVICE CO. OF COLORADO |
| To: | Berkow H Office of Nuclear Reactor Regulation |
| References | |
| P-87013, TAC-54373, NUDOCS 8707150534 | |
| Download: ML20235J400 (46) | |
Text
Public Service- ="/h.e P.o. Box 840 D"r, co 8020m 2420 W. 26th Avenue, Suite 1000, Denver, Colorado 80211 R.O. WILLIAMS, JR.
NLEAR PERAtloNS January 16, 1987 Fort St. Vrain {
Unit No. 1 P-87013 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission !
Washington, D.C. 20555 -l Attention: Mr. H. N. Berkow, Director ;
Standardization and Special l Projects Directorate Docket No. 50-267 i
SUBJECT:
Draft Fire Protection {
Program SER Comments {
REFERENCE:
- 1) NRC Letter, Heitner !
to Williams, dated !
11/18/86(G-86607) l J
Dear Mr. Berkow:
1 In accordance with the instructions contained in Ref. 1, PSC submits l herewith its comments on the draft Fire Protection Safety Evaluation j Report for Fort St. Vrain, in Attachment 1 to this letter. !
If you have ary questions concerning this matter, please contact Mr.
M. H. Holmes at (303) 480-6960.
Very truly yours, R.0. Williams, Jr.
Vice President Nuclear Operations '
l R0W/RS:jmt i 1
Attachment j G51$ $k0bb$ ,
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4 u____________________________________-____________ __ _ _ _ _ - - - - - - - . - -__ __________J
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i ATTACHMENT 1 TO P-87013 PSC COMMENTS ON ORAFT FIRE PROTECTION SAFETY EVALUATION FOR FORT ST. VRAIN l
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'l Attachment 1 to P-87013 Page 2 TABLE =OF CONTENTS Section Page ,
NRC November 18, 1986 Letter..................................... 1-1.0 Introduction ................................................ 10 ,
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1.1 Safe Shutdown Systems....................................... 12-2.1 Safe Shutdown Systems....................................... 13 2.2 Exemption Request for 7 Free Room Control Complex and Di esel Generator 'bom. . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 2.3 Exemption Request for Control Room . . . . . . . . . . . . . . . . . . . . . . . . . 26 2.4 Exemption Request for Turbine Building . . . . . . . . . . . . . . . . . . . . . . 27 2.5 Exemption Request for Access Control Bay ................... 30 2.6 Exemption Requests for Outside Areas- '
Exterior Routing and Turbine / Reactor Buildings - Common Wall..................................... 31 2.8 Exemption . Request for Emergency Lighting . . . . . . . . . . . . . . . . . . . 33 2.9 Exemption Request for Reactor Building . . . . . . . . . . . . . . . . . . . . . 35 2.11 Building 10................................................. 38-4.3 Open Itere ................................................. 40 References....................................................... 44
1 Attachment 1 j
' to P-87013 1 Page 3 NRC Letter, Heitner to Williams, November 18,1986(G-86607)
Page 1
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NRC STATEMENT i
". . . you are requested to provide'a status report containing the following information: !
Status of the related modifications and schedule for their completion,"
i PSC COMMENT The following table includes the status of the modifications and their schedule for completion. It is a condensed version of FSV Appendix R Evaluation, Vol. 4, Table 4-1, and supercedes all previous revisions submitted to that table.
TABLE 4-1 1
SCHEDULE OF PROPOSED MODIFICATIONS l Item Description Status 4.1 Fire Doors Modification Complete (Note 1) 4.2 Penetration Seals Modification Complete 4.3 Fire Dampers Modification Complete 4.4 Emergency Lighting To be Completed by 8/26/87 4.5 Fire Detection To be Completed by 1/23/87 4.6 ACM Backfeed Install by Startup l Following 4th Refueling l 4.7 Ventilation Damper Back-up Modification Complete i Bottle 4.8 Portable Ventilation Fans Modification Complete 1 4.9 Permanent Turbine Water Install by Startup Removal Pump Following 4th Refueling f
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Attachment 1 to P-87013 Page 4 Table 4-1 (continued) 4.10 Cable Re-routes (A) Reactor Plant Exhaust Fan Construction Complete (Note 2)
(B) Bearing Water Make-up Construction Complete (C) Surge Tank Level Construction Complete Instrumentation (D) Feedwater Flow Monitoring Construction Complete (E) Feedwater Flow Associated Construction Complete Cables (F) ACM Fuel Oil Transfer Pump Construction to be Completed by 2/28/87 (G) Helium Flow Instrument Cables Construction Complete (G.1) Helium Flow Instrument Fuses Modification Complete (H) Main Steam Temperature Modification Complete Indication (I) Bearing Water Pumps Construction Complete (J) Emergency Water Booster Putaps ' Construction Complete 4.11 Valve Operability (A) Accumulator / Actuator Install by Startup Modification Following 4th Refueling (B) Valve Operator Back-up Bottles Modification Complete (C) Local Control Valve Modification Complete (D) Access Ladders Modification Complete (E) Valve Tagging Modification Complete 4.12 Service Water Return Pump Modification Complete Note 1: " Modification Complete" indicates that all construction and testing is completed, and the components are physically operational.
Note 2: " Construction Complete" indicates field construction has l been completed and the components are ready for testing. I Estimates for completion are unavailable at this time, but will be provided at a future date.
l Attachment I to P-87013 Page 5 NRC Letter, Heitner to Williams, November 18, 1986 (G-86607)
Page 1 NRC-STATEMENT
. . . you are requested to provide a status report containing the following:information . . ,. '
Status of the revision of fire protection procedures and operator training that support compliance with Appendix R, and . . ."
PSC COMMENT The FSV Operations _ Department is currently preparing the Appendix R shutdown procedure. It is scheduled to be written, . approved, operator training completed, and fully implemented by March 25, 1987, i
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to P-87013 Page 6 NRC Letter, Heitner to Williams, November 18, 1986 (G-86607)
Page 2 I
NRC STATEMENT
. . . you are requested to provide a status report containing the following information . . . ,
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- Status of your compliance with Sections C, D, and F of Generic l Letter 86-10." l PSC COMMENT I
Section C of Generic Letter 86-10 reads, in part: )
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". . . where the licensee chooses not to seek prior NRC reviev and approval of, for example, a fire area boundary, ,
an evaluation must be performed by a fire protection j engineer (assisted by others as needed) and retained for !
future NRC audit. Evaluations of this type must be written j and organized to facilitate review by a person not involved j in the evaluation . .. All calculations supporting the )
evaluation should be available and all assumptions clearly l stated at the outset." '
PSC has requested documents from our fire protection consultants.
TERA Corp., which sustain the conclusions and analyses contained in l the five volume Appendix R Evaluation and Fire Hazards Analysis and l Evaluation of Building 10, previously submitted to the NRC. PSC's i Fire Protection Engineer will review the suitability and adequacy of this documentation based on the above guidelines. It is anticipated that all supporting documentation will be in place prior to the submittal of the Fire Protection Program Plan to the NRC in July, 1987.
Section D of Generic Letter 86-10 reads, in part: ,
"For fire protection systems the licensee should have and maintain a quality assurance program that provides assurance that the fire protection systems will be designed, fabricated, erected, tested, maintained and operated so that they will function as intended. Fire Protection systems are not " safety related" and are therefore not within the scope of Appendix B to 10 CFR Part 50, unless the licensee has committed to include these systems under the Appendix B program for the plant."
~ Attachment 1 to P-87013 ,
Page 7 l PSC's FSV Quality Assurance Program is described in Appendix B of-Rev. 4 of the Updated FSAR, and implemented through the "Q"
' Administrative Procedures, particularly Q-2, " Quality Assurance Prog ram. " Besides all safety related items and activities which come (
under the auspices of the QA Program, an enhanced quality program is applicable to . non-safety related fire protection, security, and radioactive waste packaging and shippin PSC's letters of May 7, ;
1985, from Singleton to Johnson (P-85133) g.and of April- 18, 1985, from j Gahm to Johnson (P-85157), described the enhanced Quality Program, j Administrative Procedure Q-2, Attachment Q-28, details the quality.
assurance requirements for non-safety related fire protection items ]
and activities. The eleven criteria required in Attachment Q-2B address and are commensurate with the ten criteria contained in Appendix A to Branch Technical Position APCSB 9.5-1, fire protection for the QA Program.
Section F of Generic Letter 86-10 reads, in part:
. . . each licensee should include, in the FSAR update required by 10 CFR 50.71(e) that will fall due more than 6 i months after the date of this letter, the incorporation of ;
the fire protection program that has been approved' by the !
NRC, including the fire hazards analysis and major 'i' commitments that form the basis for the fire protection program. This incorporation may be by reference to specific previous submittals and the NRC approvals where ;
appropriate."
"The licensee may alter specific features of t'ne approved i program provided (a) such changes do not otherwise involve a ,
change in a license condition or technical specification or '
result in an unreviewed safety question. . ,, and (b) such changes do not result in failure to complete the fire protection program as approved by the Commission. As with other changes implemented under 10 CFR 50.59, the licensee !
shall maintain, in auditable form, a current record of all such changes including an analysis of the effects of the change on the fire protection program, and shall make such records available to NRC inspectors upon request." j l
PSC notified the NRC of the development of its FSV Fire Protection Program Plan (FPPP) in a letter to Mr. Berkow from Mr. Williams on }
October 16, 1986 (P-86572). In this letter, PSC also committed to i submit the FPPP to the NRC along 'with the FSAR update in July, 1987.- i PSC has decided to include the FPPP in the FSAR by reference, in l accordance with the guidance given in Section F above. The FPPP is j currently under review for content, accuracy, and documentation {
support.
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Attachment 1 to P-87013 Page 8 NRC Letter, Heitner to Williams, November 18, 1986 (G-86607)
Page 2 ,
NRC STATEMENT
" Note that at this time you must continue all interim compensatory measures as described in your letters of April 1 and July 11, 1985 (P-85113 and P-85245), respectively. Relief from these measures will '
be granted when full compliance with 10 CFR Part 50, Appendix R is achieved. We may request certification concerning your compliance with 10 CFR Part 50, Appendix R, at a later time."
i PSC COMMENT INfERIM COMPENSATORY MEASURES PSC letter of April 1,1985 (P-85113) transmitted Volume 4 of the FSV i Appendix R Evaluation to the NRC. This report contained a section titled " INTERIM MEASURES" commencing on page 4-9, which described various interim measures that would be appropriately applied as a safety precaution for each modification proposed. 1 PSC letter of July 11, 1985 (P-85245) committed to an increased fire watch frequency of once every twenty minutes. Page 1 of this letter contained the sentence, "If approved, it is PSC's intention that the q fire watch be discontinued once the new detection system described in Reference 1 (P-85113} is operational." ,
This fire watch, and the other interim measures, will not be needed for safe operation of the plant, once the new fire detection system is declared operational, and the other applicable modifications are completed. To continue a specific interim measure when it is no longer needed for plant safety, despite the completion of an appropriate permanent plant modification, would be a drain on PSC resources without any resulting benefit to public health and safety.
The new fire detection system is being installed, and the modifications requiring the fire watch have either been completed or soon will be, as indicated in Table 4-1 (page 3) in this Attachment.
The interim measures described in P-85113 will be continued for the four modifications which will not be completed before startup, except as noted below. Once this new fire detection system is operable, PSC plans to reduce the inspection requirements of the roving fire watch to once per hour to cover only the Turbine Water Removal Pump and its associated cabling. Once the new Turbine Water Removal Pump motor is delivered, a repair procedure will be prepared to use the new motor to replace one of the existing motors within the necessary 15 to 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, if it were to be damaged by a fire, and the remaining roving fire watch would be eliminated.
Attachment 1 !
to P-87013 l Page 9 l 1
Thus, PSC wishes to retain the option of discontinuing the fire watch and other appropriate interim measures as soon as possible. This should be reflected and allowed by the NRC final SER.
APPENDIX R COMPLIANCE -)
The NRC SER phrasing does not acknowledge that'PSC has been required to. comply with only Section III.G and III.J of 10 CFR Part 50, Appendix R, with specified exemptions noted in the.SER. The NRC SER l phasing could be interpreted to mean that PSC must comply with all of- l Appendix R, yet the NRC has granted relief from Sections III.L and )
III.0 in a June.4, 1984 letter from P. C. Wagner to' O. R. . Lee (G- l 84176), and is granting relief.from portions of III.G and III.J by means of'this SER. The applicable FSV . Fire Protection Program criteria are Appendix A to Branch Technical Position APCSB 9.5-1, PSC - '
letter August 17, 1984 (P-84281), and Sections III.G and III.J of 10 l CFR Part 50, Appendix R, with'specified exemptions.
PSC will submit an official exemption request for the remainder of 10 CFR Part 50, Appendix R, except for Section III.G and III.J within 30 days.
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-l Attachment 1 s to P-87013 i Page 10 '
i Draft Appendix R Safety Evaluation 1.0 Introduction 1 Page 1 i
NRC STATEMENT "This Safety Evaluation addresses the compliance of the Fort St.
Vrain Nuclear Generating Station (FSVl with 10 CFR Part 50, Appendix i
R, concerning the fire protection programs for nuclear power i facilities."
PSC COMMENT i This statement may be interpreted to imply application of the entire Appendix R to FSV, when such is not the case. Therefore, PSC recommends inclusion of the following sentences:
"This Safety Evaluation addresses the compliance of the Fort St. Vrain Nuclear Generating Station (FSV) with 10 CFR 'Part 50, Appendix R, ' Sections III.G and III.J. concerning fire protection programs for nuclear power facilities. The NRC
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regulatory criteria which forms the complete fire protection i licensing basis for FSV consists of the- following three o documents: 1
-l 10 CFR Part 50, Appendix R, Sections III.G and III.J.
with exemptions approved herein Appendix A to Branch Technical Position APCSB 9.5-1 l PSC letter of August 17, 1984 from Lee to Johnson (P-84281) (Reference 14). This is also contained in j Appendix A of Reference la." J i
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to P-87013 i Page 11 ]
Draft Appendix R Safety. Evaluation 1.0 Introduction 1 Page 1 J i
NRC STATEMENT l
"This evaluation discusses both the proposed safe shutdown systems !
and the exemptions requested." ]:
l PSC COMMENT _
The phrase " safe shutdown" is used incorrectly in this statement and )
others throughout this SER. This phrase is used to denote equipment 1 at FSV that is on the safe shutdown list, and used after an 1 earthquake, tornado or a high energy line break to cooldown the i plant without fuel damage. This equipment is in many cases different l than the equipment used after a fire in either congested or non- .
congested cable areas. 1 j
The proper term should be " Fire Protection Shutdown /Cooldown."
Therefore, this phrase should be corrected in the following areas:
Section Page Paragraph 1.1 1 1 1 1.2 2 8 l 2.1 3 Ti tle,1 i 2.1.2 4 2 'l 2.1.3 9 Title,1 2.1.3.1 9 2 2.3.2 12 '3 2.3.3 13 2, 4 2.4.3 14 3 2.5.3 16 2 2.6.3 19 4 2.7.2 19 1, 3 2.7.3 20 3 2.8.2 21 -2 2.8.3 22 3 2.9.3 24 3 2.10.2 25 2, 3 2.10.3 25 3 4.1 27 Ti tle, 1 l
to P-87013 Page 12 Draft Appendix R Safety Evaluation 1,1 Safe Shutdown Systems Page' 1 NRC STATEMENT "PSC responded to these questions in their December 20, 1985 letter (Reference 4) which deferred the submittal of an analysis to justify- l the effectiveness of the proposed forced circulation cooldown models and the submittal of the fire protection program, until the fourth quarter of 1986."
PSC COMMENT In a March 14, 1986 letter (SER Reference 4), PSC committed "to develop a Fire Protection Program Plan by October 1, 1986." In an October 16, 1986, letter from Williams to Berkow (P-86572), PSC notified the NRC of the development of the FPPP, and concurred with a the NRC's Mr. Heitner and Mr. Kubicki request to " await the issuance of the Appendix R Safety Evaluation Report, review any open items and' 4 incorporate them into our submittal of the FPPP along with our FSAR )
Update in July, 1987."
This agreement was also in accordance with the statement on page 4 of NRC Generic Letter 86-10:
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"Therefore, each licensee should include, in the FSAR Update required by 10 CFR 50.71(e) that will fall due more than 6 months after the date of this letter, the incorporation of fire protection program that has been approved by the NRC, including the fire hazards analysis and major coninitments that form the basis for the fire protection program."
The FPPP will be submitted for incorporation into the FSV FSAR following final NRC approval of the fire protection program as i documented in the NRC SER.
These commitments should be reflected in the NRC Statement. PSC recommends that this statement should read:
. . . cooldown models until the fourth quarter of 1986.
The approved fire protection program plan will be submitted with the next FSAR Update, per NRC Generic Letter, 86-10 ,
pending' receipt of the Appendix R Safety Evaluation Report."
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1 Attachment 1 l to P-87013 Page 13 Draft Appendix R Safety Evaluation 2.1 Safe Shutdown Systems (Evaluation) j Page 3
'NRC STATEMENT "The licensee is rerouting some of these cables to improve
- . separation, and it is expected that the adequacy' of electrical i' l separation 'w i l l be verified during NRC inspections, after modifications are complete."
PSC COMMENT PSC is concerned that this wording permits the inspectors to make independent judgements .that further separation may be required, despite approved PSC exemption requests.
I PSC recommends'that this wording be changed to:
. . . and it is expected that the. electrical separation i specified in the exemption requests and proposed modifications will be verified during NRC inspections, after modifications are complete."
m.__.. _ . . - _ _ . _ _ _ ______.
Attachment 1 to P-87013 Page 14 Draft Appendix R Safety Evaluation 2.1.1- Congested Cable Areas Page 3 1
N.RC STATEMENT "The . criteria delineated in the regulatory guidance for fires in the congested cable areas was based on the use of Alternate Cooling.
Method'(ACM).'"
PSC COMMENT The regulatory guidance referred to in 'this sentence is probably NRC Reference la noted in Section 2.1 of the draft SER and should be referenced in this section for clarity.
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Attachment 1 to P-87013 Page 15 i
Draft Appendix R Safety Evaluation 2.1.1 Congested Cable Areas Page 3 :
NRC STATEMENT "The ACM can be placed in operation in approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and provides a source of Prestressed Concrete Reactor Vessel (PCRV) Liner Cooling Water (LCW)."
PSC COMMENT The Alternate Cooling Method performs other functions, such as depressurization, as well as establishment of liner cooling.
Operational conditions establish the maximum time limits within which these functions need to be accomplished. These actions can be physically accomplished within a much shorter time period.
To clarify this, PSC recommends this statement to be changed to:
I "Under ACM liner cooldown, the initial action is depressurization which must be initiated within approximately two hours; other actions are not required for a much longer time period (e.g., liner cooling must be ,
initiated within thirty hours), but can be initiated much sooner. The ACM provides a source of Prestressed Concrete Reactor Vessel (PCRV) Liner Cooling Water (LCW)."
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l Attachment 1 to P-87013 Page 16 Draft Appendix R Safety Evaluation 2.1.2 Noncongested Cable Areas Page 4 NRC STATEMENT "The criteria delineated in the regulatory guidance for fires in noncongested cable areas were based on the requirements contained in Section III.L of Appendix. R to 10 CFR Part 50. The limiting consequences require that, 'For any single fire in a non-congested .
cable area, means shall. be available to shutdown and cool down the l reactor in a manner such that no fuel damage occurs (i.e. maximum ,
fuel particle temperature does not exceed 2900 degrees F). There l shall be no simultaneous rupture of both a primary coolant boundary '
and the associated secondary containment boundary such that no i unmonitored radiological releases of primary coolant occur.'"
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PSC COMMENT l PSC recommends use of the following wording, for a better explanation of the departure from Section III.L of 10 CFR Part 50, Appendix R requirements:
"The criteria delineated in the regulatory requirements for fires in noncongested cable areas in light water reactors is contained in Section III.L of 10 CFR Part 50, Appendix R.
However, it was acknowledged in Reference 13 that this )
Section does not apply to FSV. '
In its place, PSC letter of August 17, 1984, (Reference 14) was prepared to define the criteria applicable to FSV. This 3 requires that, "For any single fire in a non-congested, .
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' Attachment 1 to P-87013 Page '17 Draft Appendix R Safety Evaluation 2.1.2.1 Reactivity Control Page-4 I
NRC STATEMENT "Two Wide Range Nuclear Instruments (one per Train) are utilized to monitor.the core reactivity." -)
PSC COMMENT.
To clarify this statement and provide a needed exemption, PSC recommends that this be worded:
"Two Wide Range Nuclear Instruments (one per train) can be i utilized to monitor the core reactivity if undamaged by a fire. Reference 4 (response to NRC Question 14) provides justification, similar to that accepted by the NRC for BWR's, that reactivity instrumentation need not be protected."
Further, PSC recommends that the following clarification relative to the nuclear instrumentation and use of the Reserve Shutdown System be added to the end of the second paragraph of Section 2.1.2.1:
"The post-fire shutdown procedure includes an instruction that should nuclear instrumentation be unavailable, operators are to activate use of the Reserve Shutdown System, in shutting down following a fire event.":
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Attachment 1 :
to P-87013 {
Page 18 l l
1 Draft NRC Appendix R Safety Evaluation ')
2.1.2.2 PCRV Integrity j Pages 4-5 j NRC STATEMENT
- 1. "PSC subsequently provided (in Reference 4) the results of a study which found that the absence of liner cooling had no significant effect on maximum fuel -or orifice valve v temperatures."
- 2. " Based on the results of the above study and the ability to provide liner cooling water via the ACM as discussed -in Section 2.1, we find the PCRV integrity provisions to be acceptable." .
i PSC COMMENT
- 1. This section discusses the effects of the Appendix R shutdown model for non-congested cable areas only. In this non-congested cable area model, forced circulation is used.
It has been shown in NRC Reference 4 that when forced circulation is used to cool the reactor after a 1.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> dclay, that liner cooling is not needed. During this cooldown, the ACM does not supply liner cooling. The ACM supplies liner cooling only after a congested cable area fire. The wording in this section may tend to confuse this issue. ,
PSC suggests that "while forced cooling is functioning" be added to the end of statement 1.
- 2. Since this section deals only with PCRV integrity following fires in non-congested cable areas, all mention of . liner cooling should be deleted from statement 2 since it is not applicable in this case. PSC suggests that the wording be:.
" Based on the results of the above study, we find the -
PCRV integrity provisions to be acceptable."
Attachment 1 l to P-87013 '
Page 19 i
Draft Appendix R Safety Evaluation 2.1.2.3.a.2 Core Heat Removal j Page 6
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NRC STATEMENT .;
"The effectiveness of the flow paths through the steam generators in f the proposed models was questioned in-. Reference 3, the requested l 1 analysis has not, as yet, been completed. However, based on a review l
of the available information, the conceptual designs of the flow paths are acceptable provided: (a)theaboveanalysisverifiesthe ,
effectiveness of the flow path, and (b). sufficient makeup water l capability is demonstrated. As stated in Reference 12, analyses are exploring various flow paths options."
PSC COMMENT Detailed analyses have been performed to verify the effectiveness of the FSV steam generators and the associated flow paths. That i analysis has been submitted by separate correspondence on December l 30, 1986 (P-86682 and P-86683).
As detailed in that analysis (P-86682), the fire protection models in Train A and Train B will support 72% and 78% reactor power levels respectively, with no modifications in the design configuration previously submitted to the NRC. Upon initiation of cooling water flow, initial flooding of the steam generators and establishment of cooling flow can be achieved in less than 15 minutes. Since the l period of steaming is limited to this short time period and operation in the closed loop requires minimal makeup, water inventory and 1 l
makeup capability are not a concern. ]
PSC is currently installing new vents to atmosphere on each of the !
steam generator EES discharge headers to enhance the current FSV.
shutdown capability. The design change analysis for these vents and any additional design changes needed to address water makeup capability will be submitted by PSC under 10 CFR Part 50.59 after the NRC final SER is received. This commitment supercedes that in PSC i letter P-86783 to submit these revised models by March 17, 1987.
An analysis has been performed on the revised fire protection shutdown models. These models rely on the new vents to atmosphere i during initial peak core cooling and then return to the models as currently submitted to the staff. The revised Train A and Train B models have been demonstrated to su3 port 82% and 87% reactor power levels, respectively. This analysis las- been documented and was submitted to the NRC staff on December 30,1986(P-86683).
The NRC should change their SER to reflect receipt of the results of this analysis.
Attachment 1 to P-87013 Page 20 Draft Appendix R Safety Evaluation 2.1.2.3.a.4 Core Heat Removal Page 6 NRC STATEMENT "A review of FSAR Section 14.4.2.1 indicates that 'One helium circulator can provide nearly 4.5% of rated flow through the reactor core when -operated by itself with condensate water supplied ~to this l water-turbine drive.' Based on this information, the primary flow requirement can be met."
PSC COMMEN'T l Although the SER conclusion is correct, it should not be based upon ,
I the FSAR quotation identified. The most limiting case is Train B's use of firewater through the firewater booster pump to drive the circulator's water turbine.
FSAR-Section 14.4.2.1 also notes:
"The corresponding case in which water from a boosted firewater pump is used to drive the water-turbine has .'also )
I been analyzed, and the results are shown in Figure 14.4-3. i Since the firewater supply is at lower pressure than the l condensate, the helium flow obtainable through the reactor '
l with one circulator operating is in this case slightly less
! than 3% of rated."
l l PSC letter of December 30, 1986 (P-86683),- Table 4-1, lists the l initial desired circulator helium flow as 1.4% for the Fire ;
l Protection Shutdown /Cooldown. l This should be reflected in the NRC's conclusions.
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l Attachment 1 to P-87013 Page 21 Draft Appendix R Safety Evaluation 2.1.2.3.a.4 Core Heat Removal Page 6 l
I NRC STATEMENT "The water used to drive the water turbine of the circulators discharges into the Turbine Water Drain Tank where it is removed by one of two Turbine Water Removal Pumps. The tank is common to both trains and the pumps are located approximately 5 feet apart, therefore, adequate separation is not maintained. However, the licensee has proposed to compensate for potential fire damage to both pumps by implementing a repair procedure using a portable Turbine Water Removal Pump and casualty cable. The adequacy of this procedure will be verified during future staff inspections."
PSC COMMENT PSC will permanently install a third full sized Turbine Water Removal Pump a minimum of 50 feet from the existing tank and pumps, This l pump will be used in normal plant operation and also meet the l criteria of redundant Appendix R emergency shutdown equipment. PSC will provide revisions to the model to reflect this change under 10 CFR 50.59 after the NRC final SER is received.
The NRC may wish to modify their statement to reflect this change.
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Attachment 1 l to P-87013 Page 22 j Draft Appendix R Safety Evaluation 2.1.2.3.c.2 Core Heat Removal and Secondary Heat Removal Monitoring Page 7-8 l
NRC STATEMENT .
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" Core heat removal monitoring is proposed to be accomplished by monitoring primary coolant flow in conjunction with secondary heat i removal monitoring (i.e., steam generator flow and exit temperature).
Coolant flow is detected by monitoring the differential pressure across the circulator; the secondary heat removal is detected by monitoring feedwater flow and steam generator exit temperature. When questioned on the adequacy of this design, PSC responded (Reference 4, Item 15) that if primary flow could be confirmed, heat would be transferred to the helium as it passed through the core, and that ,
monitoring steam generator flow and temperature would verify decay heat removal. It should be noted that the feedwater flow instruments have a range 0-400,000 lbs./hr. and the condensate flow available :
would be less that 1% of the range; this would make accurate flow !
measurement difficult.
To adequately monitor heat removal, it may be necessary to monitor a temperature differential (i.e., inlet and outlet temperatures) rather than the proposal to only monitor the steam generator exit for constant or decreasing temperature.
In addition, the circulator flow instruments lack the required separation and are included in the Reactor Building exemption request. The adequacy of the proposed instrumentation has been addressed in the exemption request evaluatien (see Section 2.9)."
l PSC COMMENT PSC submitted an analysis based upon using Appendix R Shutdown Train A following a 90 minute Interruption of Forced Cooling at 83% reactor thermal power, on December 30,1986(P-86683). This analysis was j based on 939 GPM condensate flow to the EES section of one loop with the helium flow adjusted to maintain 255 degrees F steam generator ;
water outlet temperature for the first five hours of the cooldown.
The results of this analysis show that a peak fuel temperature of 2888 degrees F would be attained, which is below the FSAR limit of !
2900 degrees F. j i
During this cooldown, the feedwater flow can be monitored to ensure j that adequate cooling occurs. The specified flow rate of 939 GPM is equivalent to about 467,000 lbs/hr, while the range on the feedwater :
loop flow rate in the Control Room is 0-1,250,000 lbs/hr. The 939 l GPM is equivalent to over 37% on this meter which allows accurate i flow measurement by the operator. !
Attachment 1 to P-87013 Page 23 Based on this analysis, it is not necessary to monitor firewater inlet temperature. A maximum conservative cooling water inlet temperature was assumed in the analysis and is the basis upon which the required steam generator outlet temperature was determined.
Reliance upon maintaining only the steam generator outlet temperature minimizes the number of indications that the operator must monitor, and complies with the accident analysis results.
The NRC concurs with the exemption request discussed in Section 2.9, yet seems to refer to the Reactor Building Exemption Request negatively here. The instrumentation ' listed in the Appendix R Evaluation is adequate'and sufficiently redundant to enable reliable monitoring of the core heat removal. ,
PSC requests _that the NRC correct the statements made in this section in light of the information given above.
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- Attachment 1
'to P-87013 Page 24 ,
i Draft Appen'ix d R' Safety Evaluation 2.2 Exemption Request for Three Room tomplex and Diesel Generator Rooms Page 10- J NRC STATEMENT
" Exemption Request for Three Room Complex an'd Diesel Generator Rooms" q PSC COMMENT
.The ."Three Room' Control Complex" has been referred to by various names throughout this SER, such as Control Room Complex and Three ,
Room Complex. The change in nomenclature is especially apparent-in ,I Section 2.2 and 2.3 of this SER. This should be changed to reflect the_ same designation given this area in the FSV FSAR, 1.e. "Three ]
Room Control Complex."
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Attachment 1 to P-87013 i 1
Page 25 Draft Appendix R Safety Evaluation 2.2.1.1 Discussion (Three Room Complex) !
Page 10 l
NRC STATEMENT
'"The steel columns in the control . room itself . are' enclosed by i concrete blocks. The steel columns are not an integral part of the I concrete wall from the standpoint of structural integrity or fife rating but are used for control complex floor connections." ,
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! PSC COMMENT PSC has recently investigated the steel columns identified above and -
their reaction to a fire. The results of this investigation will be ,
included in the FPPP under 10 CFR 50.59, after receipt of the NRC l final Fire Protection SER. To more correctly. explain the action of-these columns, the following wording should replace the NRC Statement. 1 above:
"The steel columns in the control room are partially encased by concrete blocks. The steel columns are not an integral part of the concrete wall from the standpoint of fire ;
rating. In the event of a fire in the Three Room Control ;
Complex, the vertical loads carried by these steel columns will be transferred to the concrete walls and down to the foundations. Thus, the integrity of the Three Room Control Complex boundary will remain intact."
i Attachment 1 l to P-87013 )
Page 26 l Draft Appendix R Safety Evaluation ,
2.3.3 Evaluation (Exemption Request for Control Room)
Page 13 NRC STATEMENT "If a serious fire developed, the existing halon fire suppression system would actuate to put out the fire or control it until the plant fire brigade arrived."
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PSC COMMENT-This sentence implies that the Halon fire suppression system for the control room is automatic, even though it isn't. To clarify - this ,
PSC recommends this be changed to read:
"If a serious fire developed, the existing Halon fire suppression system for the control room could be manually- ;
actuated to put out the fire or control it until the plant l l
fire brigade arrived." !
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Attachment 1 1 to P-87013 i Page 27 1
I Draft Appendix R Safety Evaluation .
2.4.2 Discussion (Exemption Request for Turbine Building)
Page 13-14 NRC STATEMENT "The principal fire hazards in the building consist of accumulations of lube oil, hydraulic oil, hydrogen gas, and cable insulation.
However, the locations which 'contain these (sic.) hazards are separated from the rest of the building by 3-hour fire rated walls, q are protected by automatic fire suppression systems, or both." J PSC COMMENT The second sentence includes " cable insulation" by reference, as being " separated from the rest of the building by 3-hour fire rated walls..." .Indeed, this is not true. This miswording should be corrected.
A recent review of substantiating documentation.by PSC determined l that the lube oil in the Turbine Lube Oil Storage and Reservoir Rooms is separated from the rest of the building by 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> rated walls, not the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated walls indicated in the Appendix R. Evaluation and the NRC SER section noted above. The hydrogen gas location will also be upgraded to a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> rating by July 1, 1987. In mitigation, it may be noted that these lube oil locations are protected by automatic ,
fire suppression. These changes will be reviewed under'10 CFR Part 1 50.59 and reflected in the FPPP submitted after receipt of the NRC final SER. Therefore, the term "3-hour rated fire walls" should be replaced with " adequately rated fire walls."
I 1 i to P-87013 i i
Page 28 l
Draft Appendix R Safety Evaluation I 2.4.2 Discussion (Exemption Request for Turbine Building) l Page 14 j al NRC STATEMENT "In Report No. 4, the licensee committed to modify and extend the existing fire detection system to provide area-wide coverage except 4 for the turbine generator operating floor and the upper level of the Access Control Bay."
PSC COMMENT l
FSV Appendix R Evaluation, Vol. 4, page 3-8 notes, "PSC proposed to j install a fire detection system throughout the area of the first two levels of the Turbine Building, but not on the third level (turbine generator floor) and at elevation 4846' 6" of the Access Control Bay." l l
PSC concurs with the commitments for the fire detection system l throughout the area of the first two levels of the Turbine Building, j For the Access Control Bay, however, PSC's intended interpretation is that the Reactor Building Exhaust Fans and Filter Area in the Access Control Bay (i.e., elevation 4846' 6") is to be the only area in the Access Control Bay that is to have a fire detection system .
This should be reflected in the NRC statement above.
i Attachment.1 to P-87013 Page 29 Oraft Appendix R Safety Evaluation
- 2.4.3 Evaluation (Exemption Request for Turbine Building)
Page 14 NRC STATEMENT "However, a fire detection system that meets the requirements of NFPA Standard No. 13 (sic.) will be. installed throughout every elevation of this fire area that does contain shutdown-related systems." q
-l NRC COMMENT 1
NFPA Standard No. 13 should be changed to NFPA Standard No'. 72E, because the former reference is for wet pipe sprinkler system, while the later correctly concerns fire detection systems. This is I properly referenced in SER Section 2.4.2 " Discussion."
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to P-87013 Page 30 l Draft Appendix R Safety Evaluation )
2.5.2 Discussion (Exemption Request for Access Control Bay) J Page 15 NRC STATEMENT i
"In addition, the licensee proposed to relocate cables and transfer I switches to the A-train fan so as to be located at least 35 feet away ' l from its redundant counterpart." l PSC RESPONSE 3 It is recommended that the NRC statement be revised as follows to '
avoid misinterpretation:
"In addition, the licensee proposed to relocate cables and ;
transfer switches to the Train A fan so tha t the switches i are located at least 35 feet away from its redundant Train B switch and cables are routed to each of the fans to enter from opposing directions to obtain maximum' separation from the redundant cables of the opposite train."
This reflects the fact that the Train A and Train B fans are physically separated by less than 35 feet, so that' the: cabling terminating at the fan can not maintain the 35 foot separation. .at l that point.
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to P-87013 ,
Page 31 ]
Oraft Appendix R Safety Evaluation j 2.6.2.1 Discussion (Outside Area - Exterior Routing Exemption) l Page 17 j NRC STATEMENT "An HVAC switchgear enclosure associated with ACM is also located south of the Turbine Building, approximately 15 feet from the building; but over 30 feet from the nearest fire protection shutdown components within the Turbine Building."
PSC COMMENT
)
i PSC measured the distances involved and recommends correcting this ]
statement to read:
i "An HVAC switchgear enclosure associated with ACM is also located south of the Turbine Building, approximately 7.5 feet from the building; but about 30 feet from the nearest i fire protection shutdown /cooldown components within the l Turbine Building."
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. Attachment 1-to P-87013 s Page 32-Draft Appendix R Safety Evaluation 1
-2.6.3 Evaluation (Exem) tion Request for
-Turbine / Reactor Buildings Common Wall)'
Page 19
.NRC STATEMENT The last line reads:
"and maintained with undamaged systems in the other five.
(sic) areas."
PSC COMMENT This should be corrected to read:
"and. maintained with undamaged system in the other fire areas." j 4
Attachment 1 to P-87013 Page 33 I
Draft Appendix R Safety Evaluation J
-2.8.2 Discussion (Exemption Request for Emergency Lighting)
Page 21 NRC STATEMENT i "2. Separate and independent power feeds for each zone covered."
PSC RESPONSE 1
This sentence.should read: l i
"2. Separate and individually protected power' feeds from l each zone covered." l Although these seven attributes were taken from page 3-20 of the FSV Appendix R Evaluation, Vol. 4, the phrase " independent power feeds" implies multiple power sources, whereas only the ACM diesel will be ;
supplying the electrical power for the emergency lighting system, i This is reflected in attribute 3 in this SER section. This should be 1 clarified by the words added above. PSC will revise this listing in j the FPPP under 10 CFR 50.59 after receipt of the NRC final Fire. I Protection SER.
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Attachment 1 ,
to P-87013 ;
Page 34 l Draft Appendix R Safety Evaluation-2.8.2 Discussion (Exemption Request for Emergency Lighting) -l Page 21 NRC STATEMENT
~" Essential valves or equipment components requiring manual operator i action (s) would be covered by local zone lighting plu: spot beams from adjacent zones."
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- PSC COMMENT I
This sentence should read: '
" Essential valve opera tors , or equipment components requiring manual operator action (s) will be covered by local zone lighting and/or spot beams."
Although this sentence was taken directly from page 3-20 of FSV-Appendix R Evaluation, Vol. 4, the deletion of "from adjacent zones"' 3
.is necessary, to avoid the implication.that the plant configuration l may permit adequate lighting from spot beams located greater than 30 ft, from the component. This also reflects the flexibility _used in the design of this emergency lighting systems. PSC will revise -this.
area in the FPPP under 10 CFR 50.59, after receipt of the final NRC Fire Protection SER.
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i Attachment I to P-87013 i Page 35 Draft Appendix R Safety Evaluation 2.9.2 Discussion (Exemption Request for Reactor Building)
Page 23 NRC STATEMENT l
"In Appendix R Compliance Report No. 4, the licensee committed to 1 install a fire detection system to provide area-wide coverage on ]
every elevation of the Reactor Building below the refueling floor." l PSC COMMENT ;
)
Page 3-25 of the PSC FSV Appendix R Evaluation, Vol. 4, comits to installing "a fire detection system throughout the area." It also notes that all of the floors have large construction openings or are {
constructed of steel grate. Not every elevation necessarily has a j specific fire detector, but the area is thoroughly covered. !
Therefore, this sentence should read: ,
l "In the Appendix R Evaluation, Report No. 4, page 3-25, the licensee committed to install a fire detection system to l provide area-wide coverage for the Reactor Building belew -
the refuelina floor." l i
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to P-87013 Page 36 Draft Appendix R Safety Evaluation 2.9.2 Discussion (Exemption Request for Reactor Building)
Page 23 NRC STATEMENT "The licensee also committed to reroute certain shutdown-related cables to achieve at least 50 feet of horizontal separation or 30 feet of separation if an intervening floor exists between redundant systems."
PSC COMMENT Certain modifications will approach these generalized criteria; however, certain modifications will not. This was explairied in Section 3.10 of the FSV Appendix R Evaluation. PSC recommends that this statement be changed as follows:
"The licensee committed to reroute certain shutdown-related cables to achieve at least 50 feet of horizontal separation or 30 feet of separation if an intervening floor exists I between redundant systems, where possible. Exceptions to {
these separation requirements of Section III.G of 10 CFR ]
Part 50, Appendix R, have been identified and evaluated on a j case by case basis and are reflected in. Table 3.10-1 of j Appendix R Evaluation, Vol 4, (Reference 1.d). The licensee 1 has committed to reroute these shutdown-related cables to ]
achieve as wide a separation as possible.
Attachment 1 to P-87013 Page 37 Draft Appendix R Safety Evaluation .
2.9.2 Discussion (Exemption' Request for Reactor Building). 1' Page 23 NRC STATEMENT j
"A portable turbine water removal pump will also be provided to compensate for the potential loss of_ redundant turbine water removal pumps on elevation 4740 ft., 6 inches."
PSC COMMENT This sentence may be revised to read:
"A th'i rd , full . sized, Turbine Water' Removal Pump, pennanently installed a minimum of 50 feet from the existing tank and pumps,.will be provided to compensate . . ." :
1 See PSC Comnent on section 2.1.2.3.a.4. ,
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to P-87013 Page 38 Draft Appendix R Safety Evaluation 2.11 Building 10 Page 26 NRC STATEMENT-
"The licensee has indicated that the guidelines aertaining to the provision of a standpipe system, yard: hydrant, fire lose, and related equipment are not applicable to Building 10. We are concerned that in the event of a fire in those areas not protected by.'an automatic suppression system, the licensee will_not have a readily available (I l
means to apply water from hose streams onto the ~ fire. We .will require that the licensee comply with Section'C.4.h (sic.).of BTP APCSB 9.5-1."
PSC COMMENT In the - event of the failure 'o f the Halon_ suppression system in Building 10, hoses from the hose stations in the Turbine Building areas can be used to apply water onto a fire occurring in Building 10 by running them through the ihree Room Control Complex. Also, a yard hydrant is located within sufficient distance of Building 10 to permit delivery of hose streams to fire occurring in Building 10.
A special "T" test performed on January 5,1987, that is available for NRC review, demonstrated that the BTP ASB 9.5-1 maximum allowable 100 feet of hose from these hose stations can cover every location in Building 10 to within 30 feet of the hose-nozzle. FSV fire fighting strategy will reflect the use of this backup method. .l Therefore, PSC believes that standpipes inside Building 10 are not required as a sufficient level of fire protection .i s afforded by.
existing hose stations and fire hydrants.
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Attachment 1 I to P-87013 Page 39
]
Draft Appendix R Safety Evaluation 2.11 Building 10 Page 2F27 NRC STATEMENT I l
"In the trip report dated September 12, 1983, which documented the ]
results of our site audit, we stated that the licensee did not have j within its organization or as a consultant a qualified fire l for the formulation protection engineer responsible and implementation of the fire protection program." q l "In Report No. 5, the licensee indicated that they will be relying upon utility staff members who are ' knowledgeable in fire protection j system operations and fire fighting techniques.' The licensee also stated that they are 'taking action to procure- qualified fire jl protection engineering services for future use.' However, the j licensee has not demonstrated that utility' staff members are so 1 qualified to meet our guidelines. Nor has the licensee provided us- I with reasonable assurance that an outside fire protection consultant I
will be 'on call' for fire protection program reviews. We will <
require that the licensee meet the guidelines contained in Section A.1 of BTP APCSB 9.5-1."
PSC COMMENT 1
In PSC letter, P-86572, dated October 16, 1986, PSC informed the NRC l of the addition of a qualified Fire Protection Engineer to the staff. ]
He is responsible for the development of the Fire Protection Program 1 Plan and is the chairman of the Fire Protection Task Force. l This section should be revised to indicate that PSC's Fire Protection Program is under the charge of a qualified Fire Protection Engineer.
The open item in Section 4.3 of this SER' associated with this issue, should be considered closed.
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Attachment 1 to P-87013 Page 40 Draft Appendix R Safety Evaluation i 4.3 0)en Items Page 23 l
l I NRC STATEMENT l l
"The licensee should provide additional information concerning the following open items:
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Effectiveness of 'the Flow Paths Through the- Steam. !
Generators,-(Section 2.1.2.3(a)(2) - PSC should ' submit the results of the analyses di.scussed in References 3 and 12;..
PSC COMMENT See PSC Response to Section 2.1.2.3(a)(2).
The NRC should modify this to reflect that these analyses have been submitted; this item should be closed and deleted from the final SER. j l
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Attachment I to P-87013' Page 41-Draft Appendix R Evaluation 4.3 03en Items-Page ZE3 NRC STATEMENT
" Adequacy of Core Heat Removal Monitoring, (Section 2.1.2.3(c)(2)).-
PSC should provide an evaluation of the potential need to monitor
' differential. temperatures .(i.e., inlet and outlet) rather than the proposal to only monitor the steam generator exit for constant or decreasing temperature."
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l PSC COMMENT.
The NRC should . consider this item closed, as a result of the discussion provided in PSC, Comment on Section 2.1.2.3(c)(2), and this item should be deleted in the final SER.
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-i Attachment 1 1 to P-87013 j Page 42
]
Draft Appendix R Safety Evaluation 4.3 Open Items ,
Page 29 NRC STATEMENT- )
\
" Building 10 Standpipe S l Equipment, (Section 2.11)ystem,
- PSC should Yard provideHydrants, Fireto Hese, a commitment comply" and Related with Section C.4.h (sic.) of BTP APCSB 9.5-1 for Building 10: .. . .
t PSC COMMENT J PSC has described information on the ability of hose stations in the Turbine Building and a yard hydrant to reach all areas of Building 10, in case of a failure of the halon suppression system, in PSC.
Comment on Section 2.11. !
Therefore, PSC does not believe that standpipes are required in l Building 10. This open item should be closed and deleted from the J Final SER. 1 I
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Attachment 1
-to P-87013 l Page 43 ]
l Draft Appendix R Safety Evaluation 4.3 n Items Page NRC STATEMENT ,
)
" Fire Protection- Staff Qualifications, (Section 2.11) - PSC should l provide a commitment to comply with Section A.1 to BTP APCSB 9.5-1." J PSC COMMENT PSC believes that the addition of a qualified Fire Protection Engineer to the PSC staff,-and his assigned duties, discussed in PSC Comment on 2.11, meets the requirements of Section A.I.to BTP APCSB q 9.5-1.
Therefore, this open item should be closed and deleted when the final SER is issued.
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Attachment 1 to P-87013 Page 44 l
Draft Appendix R Safety Evaluation References Page 30 NRC STATEMENT .,
"1. Appendix R Evaluation:
- a. Report No.1 Shutdown.model, November 16, 1984
- b. Report No. 2, Electrical Reviews, December 17, 1984 1
- c. Report No. 3, Fire-Protection, January 17, 1985 l
- d. Report No. 4. Exemptions and Modifications, April 1,1985. l l
- 2. Fire Hazards Analysis and Evaluation of Building 10 to the-BTP 9.5-1 Appendix A guidelines, Report No. 5, May 31, 1985."
PSC COMMENT The dates given above represent the initial submittal dates of the ;
.various . Appendix. R evaluation reports. However,. following the !
l initial submittal, PSC supplemented these with revised pages on )
various dates, including the SER's References 5 and 6. The revisions I to these reports also included the following letters which were not l referenced by the SER: )
PSC letter, Lee to Johnson, dated' August 30,1985(P-85301)
PSC letter, Lee to Hunter, dated September 26,1985(P-85341) l i
! The complete FSV Appendix R Evaluation set (5 Reports) should have revisions in them from these letters to the following revision and date:
Report No. 1 April, 1986 (Rev.6 Report No. 2 March, 1986 (Rev. 4 Report No. 3 April, 1986 Rev. 4 .
l Report No. 4 March, 1986 Rev. 2) l Report No. 5 August, 1985 Rev. 1) .
1 PSC recommends that these letters be included in the list of references, and that the latest revision number and date be included with the report to ensure accuracy of the reports.
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' Attachment 1 to P-87013' Page 45 Draft Appendix' R Safety Evaluation References Page 30 PSC COMMENT The following reference should be added as a result of PSC comments on Section 2.1.2, " Congested Cable Areas," page 4 (page 12 'of '.this Attachment):
- 13. NRC letter, Wagner to Lee, dated June 4, 1984-(G-84176).
- 14. PSC letter, Lee to . Johnson, dated August 17, 1984 (P-84281).
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