ML19259A853
| ML19259A853 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 12/29/1978 |
| From: | Counsil W CONNECTICUT YANKEE ATOMIC POWER CO. |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| References | |
| TASK-03-12, TASK-3-12, TASK-RR NUDOCS 7901110081 | |
| Download: ML19259A853 (9) | |
Text
'
e CONNECTICUT YANKEE ATOMIC POWER COMPANY BERLIN, CO N N E CTIC U T P. O. BOX 2 70 H ARTFORD. CONNECTICUT 06105 Tatsee oais 2 0 3 fa 6 6-6 9 t1 December 29, 1978 Docket No. 50-213 Director of Nuclear Reactor Regulation Attn:
Mr. D. L. Ziemann, Chief Operating Reactors Branch #2 U. S. Nuclear Regulatory Commission Washington, D. C.
20555
References:
(1)
W. G. Counsil letters to D. L. Ziemann dated March 6, 1978 and July 27, 1978.
(2)
D. L. Ziemann letter to W. G. Counsil dated November 9, 1978.
Gentlemen:
Haddam Neck Plant Electrical Equipment Environmental Qualification (EEQ)
In Reference (1), Connecticut Yankee Atomic Power Company (CYAPCO) submitted information on electrical EEQ to the NRC Staff. This in-formation was compiled and reformated into a standard listing by the NRC Staf f and transmitted by Reference (2) to CYAPCO for review.
As requested, CYAPC0 has reviewed the NRC Staff compilation. As a result of that review, a number of changes have been made; these changes are underlined on the attached listing.
Very truly yours, CONhECTICUT Y E ATOMICsPOWER COMPANY 9'
W W. G. Counsil Vice President Attachment
'79011100 M
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DOCKET NO. 50-213 ATTACHMENT HADDAM NECK PLANT SYSTEMATIC EVALUATION PROGRAM (SEP)
ELECTRICAL EQUIPMENT ENVIRONMENTAL QUALIFICATION DECEMBER, 1978
HADDAM NECK SEP SUBMITTAL REFERENCES (1) March 6,1978 letter, D. C. Switzer to Director of Nuclear Reactor Regulation.
(2) March 29, 1978 letter, D. C. Switzer to Director of Nuclear Reactor Regulation.
(3) July 27, 1978 letter, W. G. Counsil to Director of Nuclear Reactor Regula tion.
(4) October 10, 1978 letter, W. G. Counsil to Director of Nuclear Reactor Regulation.
(5) February 2,1978 letter, D. C. Switzer to Director of Nuclear Reactor Regulation.
(6) February 10, 1978 letter, D. C. Switzer to Director of Nuclear Reactor Regulation.
(7) December 22, 1978 letter, W. G. Counsil to Director of Nuclear Reactor Regula tion.
(8) June 21, 1978 letter, R. H. Graves to Director, Inspection and Enforce-ment, Region I.
Haddam Neck Location Documentation
References:
1.
Feb. 1, 1978 and Feb. 2, 1978, Crane Letters 2.
NUSCO Evaluations Dated 7/20/78 and 7/21/78 3.
FIRL F-C2232-01 4.
Jan 31, 1978 Letter from Limitorque 5.
Limitorque Test Report #600198 6.
FIRL F-C3441 7.
April 5, 1978, Westinghouse Letter CYW-78-518 8.
Southern Cal. Edison Report to NRC Dated February 24, 1978 9.
Collyer Technical Report #67-2 10.
IEEE Paper, R. Blodgett & R. G. Fisher, May, 1969 11.
NUSCO Evaluation Dated 7/24/78 12.
Okonite Research Report #467 13.
Stone and Webster Evaluation Dated 8/19/78 14.
March 23, 1978 Letter from General Cable 15.
January 13, 1967 Collyer Letter 16.
FIRL F-C2750 17.
Johnson Service Company Data Sheet V-24 18.
Kirchner & Bowman, Effects of Radiation on Materials and Components.
New York:
Reinhold (1974) 19.
PDCR #270 Technical Rev.
20.
February 2, 1978 and February 10, 1978 Letters, D. C. Switzer to A. Schwencer 21.
NUSCO Evaluation Dated 3/27/78 22.
Dec. 12, 1977 Letter, D. C. Switzer to E. G. Case 23.
NUSCO Evaluation Dated 7/7/78 24.
June 21, 1978 Letter, R. H. Graves to B. H. Grier 25.
NUSCO Evaluation Dated 8/3/78 26.
NUSCO Evaluation Dated 7/28/78 27.
W. G. Counsil Letter to D. L. Ziemann, Dated 10/10/78 28.
D. C. Switzer Letter to D. L. Ziemann, Dated 3/29/78 29.
NUSCO Evaluation Dated 8/1/78 30.
Stone & Webster Letter Dated 9/6/78 31.
NUSCO Evaluation Dated 7/28/78 32.
D. C. Switzer letter to D. L. Zie= ann, dated 3/6/78.
Page 1 eactor: Haddam Neck Systematic Evaluation Program SEP Submittal Time Environment Qual.
Equipment Type Reference Location Needed Parameter Req'd.
Qual.
Method Reference Motor Operated Valve 3-1/2 I
Long Temp.
260 F.
328 F.
Test Limitorque/
Pr.(Psia) 50 105 Test RH 100%
100%
Test Type: SMB Chem.
Yes b*'*
Tat 6
9 Size: 00 & 000 Rad.
1.9x10 2.2x10 Test Sub.
No 2.
2,6(T.R.),25 Motor Operated Valve 3-1/3 I
<30 Sec.
Temp.
260*F.
340 F.
Test 6
Limitorque/
Pr.(Psia) 50 120 Test 6
RH 100%
100%
Test 6
Type: SB Chem.
Yes b*'*8 Test 2,21 4
Size: 00 Rad.
8.8x10 2x10 Test 6
Sub.
No 3.
1,2,26-Motor Operated Valve 3-1/1 I
<30 Sec.
Temp.
260 F.
275 F.
Eval.
1 Teledyne/
Pr.(Psia) 50 55 Eval.
l_
T4 RH 100%
100%
Eval.
I 6
d b7
- RAD: 1.3x10 &l.4x10 4.
1,2,26 Notor Operated Valve 3-1/4 I
Long Temp.
260 F.
275 F.
Eval.
1, Teledyne/
Pr.(Psia) 50 55 Eval.
1 T10 RH 100%
100%
Eval.
1, d
xb E
1.
- RAD: 1.9x10 &3.1x10 6
6 Sub.
No
DBG-c2!
Page 2 Reactor: Haddam Neck Systematic Evaluation Program SEP Submittal Time Environment Qual.
Equipment Type Reference Location Needed Pa ramete r Req'd.
Qual.
Method Reference 5.
22,23,24,27--
Penetration, Elec.
3-5/2 I
Long Temp.
260 F.
286 F.
Test 23,27 Pr.(Psia) 50 55 Test 23,51 Rll 100%
100%
Test 23,27 Components Assembled Chem.
Yes S*'*6 Test 23,27 6
On Site - Reference Rad.
1.9x10 5x10 Test 23,27 Drawings FE-35A & FV-lE Sub.
No 6.
28,21 Terminal Block 3-5/1 I
Long Temp.
260 F.
296 F.
Test 28 GE/
Pr.(Psia) 50 55 Test 21 EB-25 RH 100%
100%
Test f5 (enclosed)
Chem.
Yes Sat.6 Test
((
5 Rad.
4.5x10 5x10 Test 28 Sub.
No 7.
19,20,--
Terminal Block 3-5/1 I
Long Temp.
260 F.
285 F.
Test 19 Westinghouse Pr.(Psia) 50 55 Test 15 805432 RH 100%
100%
Test
[5 6
8*'*7 Eval.
15 (exposed)
Chem.
Yes Rad.
1.9x10 2x10 Eval.
20 Sub.
No 8.
9,10,11 Cable 3-2/4 I
<30 Sec.
Temp.
260 F.
260 F.
Test 9
Collyer/
Pr.(Psia) 50 36 Eval.
9,11 PE/PVC RH 1001 100%
Eval.
9 Chem.
Yes Sat.6 Eval.
11 D
TO Rad.
1.3x10 5x10 Eval.
~~
Sub.
No
s DBG-c25 Page 3 Reactor: Ifaddam Neck Systematic Evaluation Program SEP Submittal Time Envi rotunent Qual.
E uipment Type Reference Location Needed Parameter R e<1' d.
Qual.
Method Reference 3
9.
10,11,12 Cable 3-3/1 I
Long Temp.
260 F.
286 F.
Eval.
12 Okonite/
Pr.(Psia) 50 65 Eval.
l_2 Butyl /PVC Ril 100%
100%
Eval.
l_2_
Chem.
Yes Sat.6 Eval.
11 6
Rad.
3.1x10 5x10 Eval.
10 Suls.
No 10.
10,11,13,2_9,30 Cable 3-3/2 I
2.5 Sec.
Temp.
260,F.
265,F.
Eval.
29 30 3
Samuel Moore /
Pr.(Psia) 50 Eval.
11_
PVC/PVC Rif 100%
Eval.
11_
Dekoron 1852 Chem.
Yes F. val.
H 6
' '6 Rad.
3.1x10 5x10 Eval.
10 Sub.
No 11.
13,14,23 Cable 3-3/3 I
Long Temp.
260 F.
302 F.
Eval.
14 General Cable /
Pr.(Psia) 50 5000 Eval.
13 HI Rll 100%
100%
Eval.
23 Rd x 10
xb Sub.
No 12.
,15,16 Cable 3-4/1 I
Long Temp.
260 F.
260 F.
Eval.
i5g Collyer/
Pr.(Psia) 50 66 Eval.
16 Silicone Rubber /
Rif 100%
100%
Eval.
l_6 Ilypalon Chem.
Yes Sat.4 Eval.
16 6
Rad.
3.1x10 5x10 Eval.
16 Sub.
No
DBG-c25 Page 4 Reactor: Haddam Neck Systematic Evaluation Program SEP Submittal Time Environment Qual.
Equipment Type Reference Location Needed Parameter Reg'd.
Qual.
Method Reference 13.
8 Transmitter, Pressure 3-2/2 I
2.5 Sec.
Temp.
260 F.
244 F.
Test Foxboro/
Pr.(Psia) 52 75 Test 8
611GM-DS1 RH 100%
100%
Test g
Chem.
N
- 6 6
Rad.
1.8x10 lx10 Test Sub.
No 14.
8 Transmitter, Level 3-2/3 I
2.5 Sec.
Temp.
260 F.
294 F.
Test 8
Foxboro/
Pr.(Psia) 50 75 Test 8_
6131[M-IISI-F RH 100%
100%
Test 8
Chem.
No
- 6 4
Rad.
1.8x10 lx10 Test 8
Sub.
No 15.
,7,31,30-Motor, Fan 3-2/1 I
Long Temp.
260 F.
320 F.
Eval.
7 Westinghouse /
Pr.(Psia) 50 95 Eval.
7 RH 100%
100%
Eval.
7 Type: CSP Chem.
Yes 88 Eval.
]
6 Frame: 684-5S Rad.
3.1x10 2x10 Eval.
-7 Sub.
No 16.
11,17,18, 32 Ai r Solenoids 3-4/3 I
1 Sec.
Temp.
260 F.
140 F.
V. Data 17,11 Johnson Serv. Co./
Pr.(Psia) 50 45 V. Data 17,11 V-24-2 RH 100%
Yes Eval.
3_2 Chem.
Yes 5
- 6
- I*
E Rad.
1.4x10 lx10 Eval.
18 Sub.
No
e DBG-c25 t
Page 5 Reactor: lladdam Neck Systematic Evaluation Progra_m SEP Submittal Time Environment Qual.
Eqaipment Typc Reference Location Needed Parameter Reg'd.
Qual.
Method Reference 17.
Junct ion ilox 2-2 Stect Box No Man. Listed, no Part //
See Item 6
Marvin' r.'Idwis'
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C 6504Bradford-Terrace Phila..PA 19149 l
1-2-79.
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Sir:
Thank you for sending me a copy of NUREG 0498 l
Please not a that my eriginal_ objections;.to-the DEIS.hr.v e not been answered..Further the information en the Radon 222 ouries and' effects are still in contradiotion to the_Testinomy of Chauncey Key ord e-n;the Perkins hearing on June:8, Ranch Mine..Please1978 and.in opposition to he numbers in the Nureg en the Merton see-those referencer fer' specific errers in NUREG 0498. Page. $ 23/.
Also a out off date for. effects o Radom 222 of 10 or 1000 years does not only ignore the na er health effectI 100,t is also he direetly in contradiction,to the PA law.
Marvin I. Lewis.
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P.O. BOX 41 Resident Manager Lycoming. New York 13093
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January 8, 1979 JAFP-79-019 Fir. Boyce 11. Grier
't United States Nucicar Regulatory Commission Region 1 631 Park Avenue King of Prussia, Pennsylvania 19406 3
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Reference:
Docket No. S0-333 Licensee Event Report:
78-105/03L-0
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Dear b!r. Grier:
I We have enclosed the referenced Licensee Event Report in accordance with Section 6.0 of Technical Specifications and USNRC Regulatory Guide 1.16.
If there are any questions concerning this report, please contact Ftr. W. Verne Childs at 31S-342-3840, Extension 207.
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ATTACllMENT TO LER 78-105/03L-0 Page 1 of 1
During normal operation while conducting Operations Surveillance Test F-ST-15A titled Pressure Suppression Chamber - Drywell Vacuum Breakers Opening and Closing Test, torus water level exceeded the requirements of Technical Specification Appendix A, Paragraph 3.7. A.1.b by approximately one (1) inch.
The actual volume of water contained in the torus did not change. When Operation's personnel opened the vacuum breakers as part of the referenced surveillance test, the Drywell to Torus differential pressure was reduced to zero psi. This differential pressure reduction allowed the vent pipe downcomers to refill with water resulting in an indicated lowering of torus level. Following completion of test, reestablishment of Drywell to Torus differential pressure, level returned to within the limits of Technical Specifications.
The plant staff intends to request a technical specification amendment to allow indicated level changes of this nature during surveillance testing.
NOTE:
LER 78-102/03L-0 and 7S-059/03L-0 are related events.
William J. Cahill, Jr.
- 4 Vice Prrsice.t Ccosolidated Edison Company of New York. loc.
4 Irving Place. New Ycrk, N Y 10003 Telephone (212) 460-3819 January 9, 1979 Re:
Indian Point Unit No. 2 Docket No. 50-247 Director of Nuclear Reactor Regulation ATIN:
Mr. A. Schwencer, Chief Operating Reactors Branch No, 1 Di,ision of Operating Reactors U. S. Nuclear Regulatory Ccmnission Washington, D. C.
20555
Dear Mr. Schwencer:
We hereby transmit, as Attachnent A to this letter, our response to your letter dated Novmber 28, 1978.
Should you or your staff have any further questions, please contact us.
Very truly yours, b5$5//j e h, 7, ?.
William J. Cahill, Jr.
attach.
Vice President I
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e ATTACHME!!T A Consolidated Edison Company of liew York, Inc.
Indian Point Unit !!o. 2 Docket :;o. 50-247 January, 1979
k A.
Containment Purging During Nornal Plant Operation-1 i
The Indian Point Unit No. -2 Containment Purge Systs and the associated containment isolation provisions are described in Sections 5.2.2 and 5.3.2 of the FSAR. The redundant 36-inch containment isolation valves in both the purge supply and exluust ducts are normally main,tained in the closed position during reactor power operation.
The Containment Purge Systs is used for containment atmosphere cleanup, cooldown and ventilation imediately prior to and during shutdown nodes when containment personnel access is required.
In addition, the purge system may be utilized to facilitate containment personnel accesn at those infrequent instances when containment entry during reactor power operation may be necessary.
Ibwever, your Novmber 28, 1978 letter has defind " limited purging" as not trore than 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> purging per year when the reactor is in other than a cold shutdown condition. Since this limitation is impractical and represents an unreasonable restriction on plant operations, we must select option (3) of your letter and, accordingly, plan to justify "unlimitcd purging". As requird by option (3), we will conduct a detailed evaluation of the Containment Purge Systs aM will respond to the issues contained in Standard Review Plan 6.2.4, Revision 1, and the associated Branch Technical Position CSB 6-4 by my 1,1979.
During reactor power operation, there is a need to provide periodic containment pressure relief to cmpensate for air leakage into containment frcm various instrument air and weld channel and containment pressurization syst s sources. As described in FSAR Sections 5.2.2 and 5.3.2, the 10-inch Containment Pressure Relief Line, not the Containment Purge System, is utilized to periodically relieve containment pressure buildup during reactor power operation. tbte that this line simply provides a pressure relief capability and does not incorporate the nomal ventilation functions of fresh air intake and air circulation that the Containment Purge Systm does.
The acceptability of pressure relief during power operation has also been deccmented in a number of other docketed references.
Applicable discussions are contained in the responses to Indian Point Unit No. 3 FSAR questions Q 11.2 and O 5.14.
In addition, present plant operating conditions are consistent with those conditions considerM by the Ccumission in IUREG-0017 (April,1976). Furthernere, the 10CFR50, Appendix I, evaluation for Indian Point Unit Ib. 2, subnitted in March, 1977, assumed continuous pressure relieving in the offsite dose calculations.
Finally, we have reviewed the instrumentation and control circuitry for the containment pressure relief line isolation valves as well as for the containment purge syst s isolation valves. As stated in the Indian Point Unit No. 2 FSAR, these valves are actuated to the closM position autmatically upon a containment isolation signal or a containment high radiation signal. Manual bypass of either signal does not affect the availability or operation of the other signal. The events at Millstone Unit 2 and Salem Unit 1 described in your Novctrbar 28, 1978 letter cannot occur at Indian Point Unit No. 2 with the present electrical design.
A-1
~
In conclusion, it is our position that present contaiment purging a d pressure relieving practices are documented and have been reviewed and accepted, and In change or new operational ccmnitments are necessary at this time.
B.
Safety Actuation Electrical Circuitry Manual Override Capability Review:
In addition to the review of safety-related electrical systans conducted at the FSAR review stage for IMian Point Unit No. 2, a number of detailed rereviews were conducted just prior to and during the first refueling outage of the unit (1976). These rereviews were conducted as part of the required subnerged electrical cmponent study, the reevaluation of the Containment Isolation Systen design and testing provisions (Appendix J), and the electrical single failure review to satisfy 10 GR 50.46 and Appendix K to 10 TR 50.
As a result of these studies, a number of hardware and software changes were effected to upgrade various features of the Indian Point Unit No. 2 safety-related circuitry.
Your Novar.ber 28, 1978 letter has requested another rereview of safety-related electrical circuitry based on new criteria and assumptions. Accordingly, we have initiatal an in-depth review to fully respond to your request. However, since this effort requires an extensive, multi-discipline review of all Indian Point Unit No.
2 safety-related electrical circuits and plant cperations, it cannot be cmpleted within a 30-day period. Upon ca:pletion of our prelinunary review and establishment of the overall scope of the review effort, we will provide a schedule for the carpletion of our review and s'dnittal of the results. We will periodically inform your Project M1 nager as to the status of the preliminary review until our for:ral schedule is subnitted.
'Ib date, we have not found any non-conforming circuits and, as stated earlier, have already reviewed the containment purge systen and containment pressure relief line control circuits and found than acceptable. Should the further checks you have requested uncover any non-conforming circuitry, we will inmediately take whatever action is necessary to prevent the potential develo; rent of an unsafe or unanalyzed condition. Any such itaas will be includcd in a future supplementary response to your Noverber 28, 1978 letter.
Based on the extensive reviews and rereviews of the Indian Point Unit No. 2 safety-related electrical circuitry conducted in the past and the upgrading of plant operations based on those reviews, we believe that operation of a bypass will affect no safety functions other than tirse analyzed and discussed on our dockets. Accordingly, no revisions to our current ad:ninistrative controls appear necessary to meet the requirenents of your Novanter 28, 1978 letter.
A-2
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.S NSF NORTHERN 5TATES POWER COMPANY M I N N E A f* C U S. M I N N C S OTA SS401 January 5, 1979 Director of Nuclear Reactor Regulation U S Nuclear Regulatory Commission Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 5 0-282 License Nos. DPR-42 50-306 DPR-60 Containment Purging During Nornal Plant Operation In a letter dated Nove=ber 29, 1978 f rom Mr A Schwencer, Chief, Operating Reactors Branch #1, Division of Operating Reactors, USNRC, we were informed of a number of events occurring at other f acilities which reduced the ef fectiveness of automatic isolation of containment purge valves. This letter requested that we respond within 30 days with our plans for dealing with this concern at the Prairie Island Nuclear Genrating Plant. Specifically, we were requested to:
a.
Review the design of safety actuation signal circuits which incorporate a manual override feature to ensure that overriding of one safety actuation signal does not also cause the typass of any othe safety actuation signal, that sufficient physical features are provided to f acilitate adequate administrative centrols, and that the use of each such manual override is annunciated at the system level for every system impacted.
b.
Agree to one of the following courses of action with respect to containment purging:
1.
Agree to cease purging, or 2.
Agree to limit purging to less than 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> / year.
Provide proof of ability of the containnent isolation valves in the purge path to close under postulated accident conditions, or 3.
Provide the basis for continued unlimited purging and a schedule for demonstrating purge isolation valves can close under postulated accident conditions, an evaluation of the impact of purging during operation on ECCS perfor-mance, an evaluation of the radiological consequences of any design basis accident requiring containment isola tion during purging, and an evaluation of containment purge and isolation instrumentation and control circuit designs.
790111oog
NORTHERN STATES POWER COMPANY Director of Nuclear Reactor Regulation Page 2 January 5, 1979 Review of Safety Actuation Signal Circuits All emergency safety features equipment control circuits have been reviewed for manual override features. The results of this review indicated that no individual equipment control circuits contain overrides of any safety actuation inputs other than local / remote selectars. These selectors are provided on certain equipment to provide for shutdown from outside the control room.
Selection of local control is individually annunciated in the control room to assure appropriate adminis trative control.
Safety actuation systems reviewed included Reactor Protection, Safeguards, and Radiation honitoring. The Safeguards System includes safety injection, containment isolation, containment ventilation isolation, contalement spray, feedwater isolation, and main steam isolation functions. Manual override of any of these systems, their subsystems, or individual circuits can only be accomplished by placing the circuitry in the test mode utilizing built-in test circuitry. Test circuit actuation only affects one of the redundant trains or channels within the system; it does not impair operability of its redundant counterpart nor does it affect operation of the other two systems. Test circuit use is annunciated in the coutrol room at least at the individual train or channel level. This facili-tates administrative control.
Containment Purging Three systems are used at Prairie Island for containment venting and purging:
a.
Contai nment purge system utilizing 36-inch supply and exhaust b.
Containment inservice purge system utilizing 18-inch supply and exhaust System (a) is a high volume purge and ventilation system (33,000 cfm) used to ventilate containment following reactor shutdown to permit access for inspection and maintenance. Two containment isolation valves are provided on each supply and exhaust line which receive an automatic closure signal on receipt of a safety injection (SI) or high radiation signal. These valves nre of the butterfly type designed to close against calculated peak accident pressure in 3 seconds.
System (b) is a low volume (4,000 cfm) purge system which provides charcoal absorption and particulate filtration of containment air prior tc release. This system is used to assist the internal cleanup system ir 1rmitting containment access when airborne radioactivity levels preclude ent.
It may also be used as a low volume normal purge and vent system. Two containment isolation valves are
r NORTHERN STATES POWER COMPANY Director of Nuclear Reactor Regulation Page 3 January 5, 1979 provided on each supply and exhaust line which receive an automatic closure signal on receipt of an SI or high radiation signal. These valves are of the butterfly type designed to close against calculated peak accidert pressure in 3 seconds.
A 90 hour0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> / year limitation on operation of either the containment purge system or the containment in-service purge system when the reactor coolant system is above 200 F could severely af fect plant availability if more than the average number of containment entries are required for inspection and maintenance. For this reason we have contacted the Westinghouse Electric Corporation with a request for technical assistance to evaluate the impact of purging during operation on ECCS performance. As noted above, the purge valves can, by design, close r;pidly unde r pos tulated accident conditions. The radiological consequences of an accident occurring during purging are believed to be inconsequential due to the rapid closdre of the purge valves. The Prairie Island containment purge and isolation instrumentation and control circuit designs are also believed to be adequate.
If, af ter discussing the requirements established by the NRC Staf f for allowing unlimited purging with Westinghouse, we find the required ECCS analysis is practical, we will provide the Commission with a schedule for submitting the necessary analysis results. This will include confirmation of acceptable radio-Logical consequences and adequate isolation instrumentation and control design.
A followup to this letter will be submitted by March 30, 1979 which will address this issue.
In the interim, use of the containment purge system and inservice purge system will continue to be minimized consistent with plant inspection and maintenance requirements and prudent plant operating practices.
Please contact us if you have additional questions related to this matter or if our schedule for conforming to your request is not satisfactory.
Yours very truly, w
Fork.
L 0 Mayer, PE Manager of Nuclear suppor Se rvices LOM/DHM/ak cc: J G Keppler G Charnoff
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,4 ye UNITED STATES
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C, 20555
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JAN 0 31979 Docket Nos.
50-373 and 50-374 Mr. Byron Lee, Jr.
Vice President Commonwealth Edison Company P. O. Box 7E7 Chicago, Illinois 60690
Dear Mr. Lee:
SUBJECT:
SECOND ROUND REQUESTS FOR ADDITIONAL INFORMATION CO LA SALLE COUNTY STATION, UNITS 1 & 2 As a result of our continuing review of the La Salle Final Safety Analysis Report, we find that we need additional information to continue our evaluation.
These questions were discussed with your staff.The specific information Please inform us after receipt of this letter of the date you can supply the requested information so that we may factor that date into our review schedule.
Please contact us if you desire any discussions or clarification of the infomation requested.
Sincerely,
?Y W fm Olan D. Parr, Chief Light Water Reactors Branch No. 3 Division of Project Management
Enclosure:
As Stated cc w/ enclosure:
Richard E. Powell, Esq.
THIS DOCUMENT CONTAINS Isham, Lincoln & Beale One First National Plaza P00R QUAUTY PAGES Suite 2400 Chicago, Illinois 60670 790111006
s 031.257 (2)
Identify the power sources for the pressure regulator and (QO31.124)
(0031.215) turbine bypass c'ontrols and describe the consequences of the failure of each source upon its loads.
031. 25 E 5:a:e :ne temoerature which was used in tne accelerated aging (0031.156)
~
(QO31.217) process.
031.259 The figures which are provided in FSAR Section 7.3 have been
( E. 2. 4. 2 :
( F7. 3-11 ;
upda:e: and mocified since Oues-ion 031.235 was issued (after (QO31.160)
(QO31.235)
Amenc.ent 35). However, the corrections seem to be incomplete.
Please clarify the discrepancy between FSAR Section 6.2.4.2.3 and FSAR Figure 7.3-11 with regard to the normal position of the LPCI pump minimum flow valves.
031.260(RSP) The response to Question 031.237 is incomplete and, therefore, (QO31.237) unacceptable. For each Class lE circuit which must be disabled in order to conduct routine surveillance testing, please provide the following for our review:'
(1) Identify the circuit (2) Provide the design basis information which is required by Section 3 of IEEE Std 279-1971
5-031.260(RSP)
(37 Describe how the requirements of IEEE Std 279-1971 for sections (Q031.237) listed below, are satisfied when due consideration is given to Section 5 of IEEE 379-1972:
(e) 4.2 (D) 4.11 (c) 4.13 (d) 4.17 (e) 4.20 Tiscri e ;
-:-3;cratior, of the circui-is '.e-ifisc af ter esti. g has been to pletec.
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e PUBLIC SERVICE COMPANY OF COLORADO P.
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84O DENVER. COLORAD0 802OI G. K. MILLEN SENIOR VICE PRESaOENT January 4, 1979 Fort St. Vrain Unit No. 1 P-79002 Mr. Themis P. Speis Advanced Reactors Branch Division of Project Management Nuclear Regulatory Commission Washington, D.C.
20555 Docket No. 50-267
SUBJECT:
Standard Technical Specifications REF: Letter T. P. Speis to J. K. Fuller, August 26, 1978 c.3ar Mr. Speis :
We have reviewed your invitation to participate in the Standard Tech-nical Specification (STS) program as outlined in the referenced letter.
Due to the unique features of Fort St. Vrain we are not prepared at this time to commit to participation in the Standard Technical Speci-fication program. We believe that the format of our present Technica:.
Specifications is adequate and in many respects meets the intent of the Standard Technical Specification.
In light of your comments in Enclosure 3 of the referenced letter, how-ever, we have recognized some shortcomings of the existing Technical Specifications, and have noted some areas where the comments are in need of further clarification. The attached comments have been pre-pared in response to Enclos tre 3.
Based upon your review and/or accep-tance of these comments, we will proceed to revise the Technical Speci-fications and submit them for your approval.
Very trul yours,
/
k-u C. K. Millen 9
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RESPONSE TO THE NUCLEAR REGULATORY C0!ffISSION RECO)DIENDATIONS AND COMMENTS ON THE FORT ST. VRAIN TECHNICAL SPECIFICATIONS 1.
Technical Specification 2 - Definitions a.
General Comment Recommendation Add a definition of instrument set points to the definition section of the Technical Specifications.
Res pons e A definition of instrument set points will be added to the definition section of the Technical Specifications.
Recommendation Delete the definition of " Low Power Operation", Technical Specifica-tion 2.5, and in its place define "Startup".
Startup may be defined as the mode switch in the run position and the Interlock Sequence switch in the startup position. Change the Technical Specification definition 2.10 " Power Operation" to relate to the mode selection switch in the run position and the ISS switch in the low power posi-tion.
Res ponse We do not intend to delete definition 2.5, Low Power Operation, nor change definition 2.10, Power Operation. Since protective actions are sequenced as a function of reactor plant power level, defining low power and power operation in terms of instrumentation readings is appropriate. We recognize as pointed out in Reference (1) that this is not consistent with light water reactor restrictions, how-ever, we have documented in numerous safety evaluations our ability to safely shutdown from the 2% power level and have justified this power level within the design basis and characteristics of the HTGR.
Arbitrary imposition of light water reactor restrictions is unaccept-able.
2.
Technical Specification 3.0 - Safety Liedts and Limiting Safety System Settings a.
Specification 3.1 - Reactor Core Safety Limit Comment / Recommendation As presently written, the total effective integrated operating time for Figure 3.1-1 is applicable only for the transient resulting in a power to flow ratio above the curve of Figure 3.1-2.
Figure 3.1-2 is applicable for power levels above 15%. Extend Figure 3.1-2 to 0% power, with a power to flow ratio of 1.0 required in all cases be-tween 15% and 0% power level. It should be noted that it is possible
. Comment / Recommendation (continued) to operate the reactor with the circulator self-turbining at power levels in excess of 15% power and/or below with a no flow condition, since there is no power to flow scram and/or no low flow scrams for this plant.
Response
Apparently, the comment regarding extension of Figure 3.1-2 has not taken into consideration the limits imposed by Specification LCO 4.1.9.
The latter LOD, Core Region Temperature Rise, Limiting Condition for Operation, is applicable for power levels from 0% to 14%. It should be noted that it is not possible to operate the reactor at power levels of 15% power and/or belcw with a no flow condition and still fulfill the requirements of LCO 4.1.9.
Recommendation Extend Figure 3.1-1 to provide applicable operating time for those power to flow ratios of 1.22 and below. Presently, an infinite time is used for these ratios. Moreover, the safety limit as written does not clearly indicate that the total damage factor to the fuel is the sum of those individually experienced at each power to flow ratio.
Consideration may be given to establishing a safety limit based on temperature.
Res pons e Figure 3.1-1 is currently under evaluation and it is our intent to extend the curve to provide applicable operating time for those power to flow ratios of 1.22 and below and 2.5 and above.
It should be noted that the safety limit has been interpreted and is being implemented in a manner which considers the fuel damage fac-tor as the sum of those individually experienced at each power to flow ratio. Specification SL 3.1 refers to the total effective integrated operating time and we have interpreted this as the sum at each power to flow ratio.
Comment It is also noted that present in-plant instrumentation is limited with regard to resolution of the time period when the power to flow ratios exceed those of this Technical Specification. The power to flow recorder, which is the primary source of this information, is a one inch per hour chart. At present the plant relies on General Atomic Company test computer data. However, we understand that you will install brush recorders which would be triggered by power flows in excess of the allowable limits.
. Res pons e The in-plant instrumentation for power to flow is presently under evaluation and various alternatives are being considered. Improved instrumentation will be provided as necessary te rulfill Technical Specification requirements for evaluating power to flow transient conditions.
3.
Technical Specification 4.0 - Limiting Conditions for Operations a.
General Cocnent Change the wording of the last sentence on page 4.0-1 which defines the time period in which corrective action must be taken as 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, to be similar to that in BWR Standard Technical Specification 3.0.4.
Response
We believe the wording of the last sentence on page 4.0-1 establishes the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period referenced except for those LCO's in which speci-fic time limits have been established by the LCO which are supported by the basis for the particular LCO.
b.
LCO 4.1.2 - Operable Control Rod. Limiting Conditions for Operation Comment It should be noted that in accordance with this Technical Specifica-tion, the reactor may be operated without operable control rods or sufficient reactivity to ensure cold shutdown up to power levels of 2%.
Consistent with other Technical Specifications, the reactor con-trol rods should be operable for startup and power operations.
Res pons e We agree with the comment. The LCO will be changed to include require-ments for control rod operability at any time when the reactor is not in a shutdown or refueling condition.
c.
LCO 4.1.3 - Rod Sequence Comment / Recommendation In the basis section of this Technical Specification there are a num-ber of peaking ratios given for different fuel regions. However, it does not appear that there are any surveillance requirements to deter-mine these peaking ratios on any regular basis. Consideration should be given to requiring periodic determination of the peaking ratios as a surveillance requirement.
. Respons e No surveillance requirement is necessary to determine the peaking ratios contained in the basis of LCO 4.1.3.
The peaking factors con-stitute one of the three bases used for establishing any control rod withdrawal sequencing which must be approved by the Nuclear Facility Safety Committee (NFSC). Therefore, adhering to the approved rod withdrawal sequence is sufficient to assure that peaking factors are in specified ranges, d.
LCO 4.1.6 - Reserve Shutdown System Recommendation This section should be reworded to require a minimum operability of the reserve shutdown system anytime the reactor is being started up, and for power or low power operation.
Response
No changes are required since LCO 4.1.6 already specifies low power (103% to 2%) operation which limits reactor power to less than 10-3g unless the conditions of the LC0 are met.
e.
LCO 4.1.7 - Core Inlet Orifice Valves Comment At present there appears to be no corresponding surveillance require-ment regarding the orifice valve position indications.
Res ponse There is no surveillance requirement regarding orifice valve position indication. The orifice valves are provided only to control region outlet temperatures specified in this LCO. The position of the ori-fice valve is therefore dictated by temperature and is irrelavant to compliance with the LCO or the safety limits addressed in the basis of the LCO.
f.
LCO 4.2.1 - Number of Operable Circulators Recommendation This Technical Specification should be changed to require one operable circulator in each loop for any startup or critical operation. It should be noted that the Technical Specification is presently appli-cable only for power operation.
. Response No change is required to LCO 4.2.1 since, as stated in the LCO, if only one circulator is operable at any time the reactor must be placed in a low power or shutdown condition. With those requirements, reac-tor power operation with only one operable circulator is not permitted.
g.
LCO 4.2.2 Recoc=endation A requirement for the operability of the backup bearing water system should be incorporated in LC0 4.2.2, in addition to the other condi-tions required for an operable circulator.
Res pons e As indicated in the Final Safety Analysis Report the backup bearing water system is non-Class I, and no credit is taken for this system in the accident analysis. The system is available strictly as a back-up and therefore there is no basis for incorporating the system into the LCO.
Re commendation Technical Specifications LCO 4. 2. 3, 4. 2.4, 4. 2. 5, and 4. 2. 6 are all Technical Specifications which are required for power operation only.
These Technical Specifications should be evaluated to determine if they should apply also to low power or startup conditions. Technical Specifications 4.2.12, 4.2.13, and 4.2.14 are again applicable only at power operation. These should be evaluated to determine if they should also be applicable any time the reactor is to be started up or at low power conditions.
Response - LCO's 4. 2. 3, 4. 2. 4, 4. 2. 5, 4. 2. 6, 4. 3.1, 4. 3. 2, 4. 3. 4, 4.3.5, and 4.3.6 The objectives of Sections 4.2 and 4.3 of the Technical Specifications are to ensure the capability to cool the reactor core.
The requirements for operability of systems or components during power operation, where specified in individual LCO's is adequate to assure that the capability exists to fulfill the intent of the LC0 for which the statement is applicable.
The configurations required by the LCO's listed above provide suffi-cient capability for a safe shutdown of the plant and for safe shut-down cooling as described in Section 10 of the Final Safety Analysis Report.
Based upon that capability, the LCO's listed above do not need to be changed to include startup or low power operation.
Response - LCO's 4.2.12 and 4.2.18 The provisions contained in the above listed LCO's are adequate to assure that the capabilities exist to fulfill the functional require-ments stated in the bases of the LCO's and Section 9.6 of the Final Safety Analysis Report.
Based upon those capabilities, the above listed LCO's do not need to be changed to include startup or low power operation.
Response - LOO's 4.2.13 and 4.2.14 The provisions contained in the above listed LOO's are adequate to aseure that the capabilities exist to fulfill the functional require-ments stated in the bases of the LOO's and Section 5.4 of the Final Safety Analysis Report.
Based upon those capabilities, the above listed LCO's do not need to be changed to include startup or low power operation.
Recommendation The requirements for "at power" operation is contained in the major-ity of the remaining Section 4, " Limiting Condition for Operation",
Technical Specifications. Rather than identify each specific LCO with the requirements for "at power" conditions each Technical Specifica-tion should be evaluated to determine if it should be applicable for startup and low power conditions.
Response
The following Section 4 Limiting Conditions for Operation do not need to be changed to include either startup or low power conditions:
- 1) LCO 4. 2. 2, 4. 2. 7, 4.2. 8, 4. 2.9, 4. 2.10, 4. 2.11, and 4. 2.15
- 2) LOO 4. 3. 8, 4. 3.9, and 4. 3.10
- 3) LCO 4.4.3 and 4.4.5
- 4) LCO 4.5.2
- 5) LOO 4. 7.1, 4. 7. 2, 4. 7.3, and 4. 7.4
- 6) LCO 4. 8.1, 4. 8. 2, and 4. 8. 3
- 7) LCO 4.9.1 and 4.4.2
- 8) LCO 4.10.3 and 4.10.4
. Response (continued)
Since these LCO's either already include startup and low power or do not apply directly to reactor operation or are applicable regardless of reactor operational modes, no changes are required.
The startup or low power aspects of the following Section 4 Limiting Conditions for Operation are addressed elsewhere in this correspondence:
- 1) LCO 4.1.2 under item 3.b.
- 2) LCO 4.1.6 under item 3.d.
- 3) LCO 4.2.1 under item 3. f.
LCO's 4.2.16 and 4.2.17 The provisions contained in the above listed LCO's are adequate to assure that the capabilities exist to fulfill the functional require-ments stated in the bases of the LCO's.
These LOO's are applicable to a temporary system which would be uti-lized only if a permanent loss of forced circulation as a result of a major event renders normal plant electrical equipment inoperable.
Based upon the requirement for the system, the above listed LCO's do not need to be changed to include startup or low power operation.
LCO 's 4. 3. 3, 4. 3. 7, 4. 5.1, and 4. 6.1 The provisions contained in each of the above listed LCO's are ade-quate te assure that the capabilities exist to fulfill the functional requirements stated in the basis of each LCO.
Based upon the requirements for the systems, the above listed LCO's do not need to be changed to include startup or low power operation.
LCO's 4.4.2 and 4.4.6 The provisions contained in the above listed LOO's are adequate to assure that the capabilities exist to fulfill the functional require-ments stated in the bases of the LCO's as discussed in Final Safety Analysis Report Amendment No.17, Question 7.5.
LCO 4.4.4 Based upon the evaluation for the LCO's 4.2.3 through 4.2.6, the provi-sions contained in LCO 4.4.4 do not need to be changed to include startup or low power operation.
- Response (continued)
LCO's 4.10.1 and 4.10.2 The provisions contained in the above listed LCO's are adequate to assure that the capabilities exist to fulfill the functional require-ments stated in the bases of the LCO's.
With the operability of the Alternate Cooling Method System, the conditions for core heat removal under accident conditions can be fulfilled.
Based upon the requirements for the systems, and operability of the Alternate Cooling Method, the above listed LCO's do not need to be changed to include startup or low power operation.
h.
Technical Specification 4.4 - Instrument and Control Systems - Limiting Condition for Operation Comment /Re commendation This portion of the Technical Specifications does not address the loop dump system. The loop dump and control systems should either be incorporated as a part of the existing tables and/or a new and separate table defined. On all Section 4.4 Technical Specifications it would appear to be legally permissible to bypass all operable chan-nels for periods of up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Likewise, there is no minimum amount of instrumentation that must be available to the operator to support plant operations. Example: All control rod position indi-cation may be inoperable and the reactor operated at power. These latter two comments would appear to be apparent problems from the standpoint of inspection and enforcement.
Response
Although not mentioned specifically as the loop dump system, the sys-tem is addressed in LCO 4.4.1, Table 4.4-1, Item Numbers 4 and 7, and Table 4.4-2, Item Numbers 5a and Sb.
It should be noted that Speci-fication LSSS 3.3, Table 3.3-1, Parameters 2(a) and 2(b) include Steam / Water Dump Functions. Therefore, no change is required.
It is not legally permissable to bypass all operable channels for periods of up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The Permissable Bypass Conditions, if any, are stated on Tables 4.4-1, 4.4-2, 4.4-3, and 4.4-4 for each Func-tional unit heading. Note (f) of Notes for Tables 4.4-1 through 4.4-4 states that "The inoperable channel must be in the tripped condi-tion, unless the trip of the channel will cause the protective action to occur." Note (f) relates to the Minimum Operable Channels and Minimum Degree of Redundancy headings of Table 4.4-1 through 4.4-4
- Response (continued)
Both the Degree of Redundarty and an Operable Channel are defined in LCO 4.4-1.
Degree of Redundancy is defined as the difference be-tween the number of operable channels and the minimum number of oper-able channels which when tripped will cause an automatic system to trip. A typical two out of three logic scheme requires a minimum of two operable channels to achieve a minimum degree of redundancy of one.
Further, since Note (f) requires that the inoperable channel be in the tripped condition, one of the channels is operated thereby requiring only one other of the two unaffected channels to operate and initiate a two out of three logic trip to provide protective ac-tions.
1.
LCO 4.5.1 - Reactor Building Recommendation For the establishment of reactor building integrity, a condition should be added which requires that other building penetrations be secured.
In the past, it was found that piping penetrations into the reactor building were not sealed. A surveillance requirement for inspections of the Reactor Building penetrations at each refueling should also be considered.
Respons e Specification LCO 4.5.1 defines the conditions required to maintain the reactor building integrity. Since the established criteria as-sure the integrity of the reactor building, no changes are required.
- j. Technical Specification LCO XXXX - Control Room Ventilation Recommendation There is no apparent LCO or Surveillance Requirements for the control room ventilation system. This should include the filter system asso-ciated with the control room ventilation. The charcoal filters asso-ciated with the control room ventilation system have not been described in the Final Safety Analysis Report.
Response
Specification LCO 4.4.2 - Control Room Temperature, will be revised to include requirements for the control room ventilation system.
Specification SR 5.4.7 - Control Room Temperature, will be revised to include surveillance testing requirements for the control room ventilation system.
It appears that the charcoal filters associated with the control room ventilation system have not been described in the Final Safety Analy-sis Report. A description of this system can be included in the Final Safety Analysis Report at such time that the Nuclear Regulatory Commis-sion provides guidance and/or regulation for updating the Final Safety Analysis Report.
. 4.
No Number 4 - skips from 3 to 5.
5.
Technical Specification 5.0 - Surveillance Requirements a.
SR 5.2.2 - Tendon corrosion Suveillance Comment / Recommendation This surveillance requirement provides no acceptance criteria for either corrosion of the wire samples or corrosion products as found in the atmosphere of the tendon tube. The findings from this inspec-tion should be reported to the Nuclear Regulatory Commission.
Res pons e Although acceptance criteria are not listed in Specification SR 5.2.2, we have requested General Atomic Company to supply the necessary acceptance criteria which we will include in the SR 5.2.2 surveillance tests.
With approval of Amendment Number 16 to the Fort St. Vrain Technical Specifications, reporting results of Technical Specification surveil-lance tests is no longer required. Results of surveillance tests are available and are regularly reviewed by Nuclear Regulatory Commission Inspection and Enforcement.
b.
SR 5.2.3 - Tendon Load Cell Surveillance Comment This surveillance requirement provides no acceptance criteria for possible shif ts and load cell reference points. If shifts do occur, a report of these findings should be provided to the Nuclear Regula-tory Commission.
Respons e Although acceptance criteria are not listed in Specification SR 5.2.3, the surveillance test SR 5.2.3-X does include acceptance criteria.
With approval of Amendment Number 16 to the Fort St. Vrain Technical Specifications, reporting results of Technical Specification surveil-lance tests is no longer required. Results of surveillance tests are available and are regularly reviewed by Nuclear Regulatory Commission Inspection and Enforcement.
c.
SR 5.2.4 - PCRV Concrete Crack Surveillance Comment This surveillance requirement does not provide for reporting the findings of the concrete crack surveillance inspection to the Nuclear Regulatory Commission.
. Res ponse With approval of Amendment Number 16 to the Fort St. Vrain Technical Specifications, reporting results of Technical Specification surveil-lance tests is no longer required. Results of surveillance tests are available and are regularly reviewed by Nuclear Regulatory Com-edssion Inspection and Enforcement, d.
SR 5.2.5 - Liner Specimen Surveillance Comment This surveillance requirement provides no basis for acceptance cri-teria changes in notch toughness. It also does not provide for re-porting the findings from this surveillance to the Nuclear Regulatory Comnission.
Res pons e Although acceptance criteria are not listed in Specification SR 5.2.5, the surveillance test SR 5.2.5-X does include requirement for analy-sis per ASTM-E-185-70. With approval of Amendment Number 16 to the Fort St. Vrain Technical Specifications, reporting results of Techni-cal Specification surveillance tests is no longer required.
Results of surveillance tests are available and are regularly reviewed by Nuclear Regulatory Commission Inspection and Enforcement.
e.
SR 5.2.6 - Plateout Probe Surveillance Comment This specification does not provide for reporting the findings from the plateout probe surveillance to the Nuclear Regulatory Commission.
Response
With approval of Amendment Number 16 to the Fort St. Y<ain Technical Specifications, reporting results of Technical Specification surveil-lance tests is no longer required. Results of surveillance tests are available and are regularly reviewed by Nuclear Regulatory Commission Inspection and Enforcement.
f.
SR 5.2.7 - Water Turbine Drive Surveillance Recommendation Acceptance criteria should be established in this specification for operation on water turbine drive using feedwater or condensate at re-duced pressure to simulate fire discharge pressure as motive power.
The circulator should be capable of the helium flow requirements established in the Accident Analysis Chapter of the Final Safety Analysis Report.
Res ponse Although acceptance criteria are not listed in Specification SR 5.2.7, the surveillance test SR 5.2.7-A includes curves which specify ninimum acceptable speeds for operation on water turbine drive using feedwater, condensate, or condensate at reduced pressure to simulate fire water conditions. The curves cover the range of helium densities and uti-lize requirements established in Final Safety Analysis Report Section XIV, Part 14.4 to specify the acceptable speeds for each condition.
g.
SR 5.2.8 - Bearing Water Makeup Pump Surveillance Recommendation Acceptance criteria should be established for normal makeup flow and emergency makeup pump flow. Flow rates should be verified during the surveillance test.
Response
Although acceptance criteria are not listed in Specification SR 5.2.8, surveillance test SR 5.2.8a-Q, utilizes the pump curve for the normal bearing water makeup pump (P-2105) to establish the minimum operability of the pump. Surveillance test SR 5.2.8abc-Q includes relief valve operation to verify operability of the positive displacement emergency bearing water makeup pump (P-2108).
h.
SR 5.2.10 - Engine-Driven Fire Pump Surveillance Recommendation This surveillance section should be expanded to include the following:
uinimum fuel inventory; minimum acceptable pump discharge and flow rate; requirements to perform an annual inspection of the pump and diesel engines; and an annual test under full load conditions.
Response
Specification LCO 4.2.6 - Firewater Pumps, requires that r.t least 325 gallons of fuel be in storage. Verification of this requirement is performed and recorded once per shift with 600 gallons specified as the reorder point.
The weekly surveillance test is adequate to ensure proper operation of the pump and associated control.
Preventive Maintenance Inspection Procedures, PM 45.5 and PM 45.6, provide inspections of the pump and diesel engine in accordance with the manufacturer's suggested schedules.
1.
SR 5.2.13 - PCRV Concrete Helium Permeability Surveillance Comment This specification does not provide any acceptance criteria or reporting requirements of the results to the Nuclear Regulatory Commission.
Res pons e Although acceptance criteria are not listed in Specification SR 5.2.13, surveillance test SR 5.2.13-X does include expected test results.
With approval of Amendment Number 16 to the Fort St. Vrain Technical Specifications, reporting results of Technical Specification surveil-lance tests is no longer required. Results of surveillance tests are available and are regularly reviewed by Nuclear Regulatory Commission Inspection and Enforcement.
- j. SR 5.2.14 - PCRV Liner Corrosion Surveillance Requirement Comment this specification does not provide any acceptance criteria for liner thinning or reporting of the results to the Nuclear Regulatory Commis-s ion.
Response
Although acceptance criteria are not listed in Specification SR 5.2.14, surveillance test SR 5.2.14-X does include provisions for evaluating data and determining the success or failure of the test.
With approval of Amendment Number 16 to the Fort St. Vrain Technical Specifications, reporting results of Technical Specification surveil-lance tests is no longer required. Results of surveillance tests are available and are regularly reviewed by Nuclear Regulatory Commission Inspection and Enforcement.
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SR 5.2.17 - Helium Circulator Pelton Wheels
_ Comment This specification does not provide for reporting the findings of the inspection of the Pelton wheels to the Nuclear Regulatory Commis-sion.
Respons e Specification SR 5.2.17 does not provide for reporting the findings of the inspection of the Pelton wheels to the Nuclear Regulatory Commission. However, Specification 5.2.18 which is performed at the same time as SR 5.2.17 does specify submitting results of examinations to the Nuclear Regulatory Commission. Although not specifically cen-tioned in SR 5.2.17 it is our intent to include the results of the examination of the Pelton wheel in the helium circulator examination report under SR 5.2.18.
If necessary, a change to SR 5.2.17 to this effect could be made.
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SR 5.3.1 - Steam / Water Dump System Valves Surveillance Requirements Comment This specification does not establish any acceptance criteria for the opening time of the steam / water dump valves.
Res pons e Although acceptance criteria are not listed in Specification SR 5.3.1, the surveillance test SR 5.3.1-Q specifies an opening time for the steam / water dump valves, the results of which are availcble for Nuclear Regulatory Commission Inspection and Enforcement review.
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SR 5.3.2 - Main and Hot Rehest Steam Stop Check Valves, Surveillance Requirements Comment This specification does not establish any acceptance criteria for the stroking time of the main steam and hot reheat steam stop check valves.
Respons e The testing specified in Specification SR 5.3.2 is adequate to assure the functional reliability of the main and hot reheat steam stop check valves as stated in the basis of SR 5.3.2 and Section X, Part 10.5 of the Final Safety Analysis Report. It is not possible to check the stroking time of these valves during normal operation; therefore func-tional tests are performed. Acceptance criteria for the stroking
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time will be specified in surveillance test SR 5.3.2-A.
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SR 5.3.3 - Bypass and Safety Valves. Surveillance Requirements Recommendation Surveillance requirements should be established for the calibration of the instrumant and controls systems associated with the bypass and safety valves.
Res ponse Instrumentation associated with the bypass and safety valves is cali-brated on an annual basis. This frequency is adequate to ensure opera-bility of the equipment, k
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(l jjt 5.3.4 - Sa fe Shutdown Cooling Valves, Surveillance o.
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Ra:ommendation
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A table should be (stablished in this section for those valves nec-
.essary for actuation of the safe shutdown cooling code. Where appro-g<
priate, limiting conditions for operation should be established re-quiring operabilit.v of these valves.
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33urveillance test CR 5.3.4-SA contains a list of valves necessary for safe shutdown caeling.
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,Rhere appropriate, limiting conditions for operation vill be established
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requiring operability of safe shutdown cooling valves.
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SR 5.6.1 - Standhv Diesel Generator Surveillance Recoc=erdation The diesel should be loaded to 100% instead of the 50% presently in' the Technical Specifications. The interval for performance of the lo ? capacity test sho21d be changed to monthly. A surveillance
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conum.lly or at the refueling interval.
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r-Res pons e Specification SR 5.6.1 will be revised to a monthly 100% load test.
Inspection and Preventive Maintenance Procedure, PM 92.10, specifies the quarterly, semi-annual, and annual inspections of the standby diesel units in accordance with the manuf acturer's recocmendatier.s.
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LOO 4.2.10 - Loop Impurity Levels. High Tecoeratures Recommendation Some discrepancy exists between this LCO and the Final Safety Analy-sis Report (page 4.2-2). Revise the LCO and discuss the cerits if any of an inte, rated equivalent i= purity level and a maximum number 9f cycles that the amounts may be exceeded for 10 days. From the LCO it would appear to be possible to operate for 10 days with oxidant
~1mpurity levels exceeded by a factor of 10, then decrease the tempera-ture to below 1,200*F for one day and repeat the cycle ind.initely.
The staff finds this unacceptable.
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Res pons e The LCO will be changed to clarify the time and nr2ber of cycles as a function of integrated equivalent impurity levels, if applicable.
We are pursuing this matter with General Atomic Company, and will ree-oncile the apparent differences between the LCO and the Final Safety Analysis Report.
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LCO 4.6.1 - Auxiliary Electric System, Limiting Conditions for Opera-cion Comment Part d4 requires that the reactor shall not be operated at power un-less 500 gallons of fuel exists in each day tank of the diesel genera-tors. Since these tanks have a 500 gallon capacity, it would be ex-tremely difficult to maintain the set level af ter even a short inter-val of running the diesels.
Response
A change to Specification LC0 4.6.1 has been submitted to the Nuclear Regulatory Commission which requested a reduction in the fuel required to 325 gallons.