ML19296B950

Forwards Addl Info on NUREG-0578,in Response to NRC Review Team 800121-22 Site Visit

ML19296B950
Person / Time
Site:	Fort Saint Vrain
Issue date:	02/20/1980
From:	Swart F PUBLIC SERVICE CO. OF COLORADO
To:	Varga S Office of Nuclear Reactor Regulation
References
RTR-NUREG-0578, RTR-NUREG-578 P-80028, NUDOCS 8002220409
Download: ML19296B950 (15)

Text

.

public service company ce conende February 20, 1980 Fort St. Vrain Unit No. 1 P-80028 fir. Steven A. Varga Acting Assistant Director for Light Water Reactors Division of Project Management U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Docket # 50-267

Subject:

Additional Information Resulting from Three flile Island Unit 2 NRC Review Team Site Visit of January 21-22, 1980 Gentlemen:

As a result of the subject flRC Review Team Site Visit, PSC committed to supplying additional information on the following flVREG-0578 sections:

Section 2.1.2 - Performance Testing for BWR and PWR Rellef and Safety Valves Section 2.1.3.a - Direct Indication of Power Operated Relief Valve and Safety Valve Position for PWR's and BWR's Section 2.1.4 - Containment Isolation Provisions for BilR's and PWR's Section 2.1.8.a - Improved Post-Accident Sampling Capability Section 2.1.8.b - Increased Range of Radiation lionitors Section 2.1.8.c - Improved In-Plant Iodine Instrumentation Section 2.2.1.a - Shift Supervisor Responsi,bilities Section 2.2.1.c - Shift and Relief Turnover Procedures Section 2.2.2.b - Onsite Technical Support Center Section 2.2.2.c - Onsite Operational Support Center 8 002geo Q f ,

Mr. Steven A. Varga February 20, 1980 Page 2 Enclosed are the PSC submittals in response to review team questions on these sections. Should clarifications or additional information on this subject be required, please contact this office.

Very truly yours, ,.

. w & ,,, w :

2- suowy

.I Frederic E. Swart Nuclear Project Manager Enclosures

Section 2.1.2 Performance Testing for BWR and PWR Relief and Safety Valves The following question was raised by the THI-2 review team during the January 21-22, 1980 site visit.

1. Provide additional justification that PCRV relief valves are qualified for intended service.

PSC Reply:

PSC intends to rely upon the qualification testing program being performed by EPRI and will abide by applicable recommendations resulting from that program, taking credit for features unique to the Fort St.

Vrain HTGR such as the absence of two phase or liquid phase flow conditions.

Section 2.1.3.a Direct Indication of Power-0perated Relief Valve and Safety Valve Position for PWR's and BWR's The following question was raised by the TliI-2 review team during the January 21-22, 1980 site visit:

1. Licensee must present argument (in writing) why relief valve position indication, qualified per the October 30, 1979 Denton letter, is not necessary.

PSC Reply:

PSC's original reply, PSC letter P-79249 dated October 29, 1979, discussed the unique features of the gas-cooled reactor regarding the possibility of and consequences of a primary coolant overpressure accident.

PSC also described the devices (pressure and radiation alarms and pressure indicators) that would indicate, although indirectly, relief valve status. It should also be noted that the two interspace pressure alarms and the three primary coolant pressure indicators, all of which are in the control room, are all safety grade instruments. Furthermore, it should be emphasized that primary coolant pressure is the most critical parameter during any loss of coolant condition in a gas-cooled reactor, that the reactor can be adequately cooled at atmospheric pressure in the primary coolant system at decay heat levels, and that complete depressurization of the primary coolant inventory via the PCRV relief filter results in a radiation dose at the exclusion area boundary of only about 2 percent of the 10CFR100 limit.

Other factors that PSC has taken into consideration in concluding that the present relief valve instrumentation is adequate include:

1. The FSV gas-cooled reactor is distinct from PWR's and BWR's in that maintenance of primary coolant inventory and pressure is not crucial to prevention of core damage. Due to inherent design features, the FSV gas-cooled reactor does not utilize coolant make-up to protect the core in case of a loss of coolant accident.

2. Failure of a safety-relief valve to reset would be readily apparent to the operator by the circumstances of the malfunction and existing safety-grade instrumentation.

3. The only apparent operator action to mitigate the consequences of such a malfunction would be isolation of one or both of the safety relief valves. However, isolation of even one of the safety relief valves is specifically prohibited by both the FSAR (Section 6.8.2.1) and the Technical Specifications (LC0 4.2.7) except for maintenance and/or testing when the PCRV is depressurized. The pertinent ASME Code requirements also lead to the conclusion that isolation of a safety relief valve should not be allowed. In light of this and the potential for more severe consequences due to PCRV damage resulting from compromising the pressure relief train as compared with the consequences of releasing filtered primary coolant at a reasonably slow rate, it is debatable whether such action should be taken in any case.

Therefore, it is PSC's considered position that the addition of direct position indication to the safety relief valves is not necessary or justifiable for the FSV HTGR.

Section 2.1.4 Containment Isolation The following question was raised by the TitI-2 review team during the January 21-22, 1980 site visit.

1. Provide additional information to show how the current FSV

" containment" penetration isolation provisions satisfy flVREG-0578 Section 2.1.4 (1, 2 and 3) requirements. Specific information (auto / manual, isolation signal, essential /non-essential,1ine size and closed system) should be provided for each system penetrating " containment". Describe the basis for classifying systems as essential.

PSC Reply:

The following information is provided in addition to that contained in PSC letter P-79299, dated December 12, 1979.

Piping penetrating primary and/or secondary containment are listed in the attached Table 2.1.4-1. Each penetration is classified per the most recent 10CFR50, Appendix A, General Design Criteria and the isolation provisions are briefly identified.

As PSC has indicated in previous submittals, the FSV plant does not utilize a reactor containment similar to PWR and BWR reactor plants.

Therefore, the concept of essential /non-essential systems does not apply to the Fort St. Vrain HTGR as it would to water-cooled plants. Disregarding the differences for the purposes of this reply only, PSC has c:assified the piping penetrating containment as " essential" if it is required to perform an active role in providing the following safe-shutdown functions:

(1) Reactivity Control Lines servicing the control rod drives and reserve shutdown hoppers are classified " essential" for this purpose.

(2) Core Cooling Lines servicing the Helium Circulators, the Steam Generators and PCRV Liner Cooling System are classified " essential" for this purpose.

(3) Depressurization Lines servicing the Helium Purification train are classified

" essential" for this purpose.

(4) Containment Lines servicing the circulator brake and shutdown seal are classified " essential" for this purpose.

(5) Overpressure Protection Lines servicing the PCRV relief valves and penetration interspace relief valves are classified " essential" for this purpose.

This classification in no way alters or commits PSC to alter the basis upon which the plant is licensed.

These classifications have been made for the purposes of responding to specific NRC requests even though PSC does not consider these SWR and PWR classifications and their implications applicable to the FSV HTGR.

Automatic operation of isolation valves is initiated by diverse containment isolation signals as appropriate to the individual system purposes and designs.

Nonessential systems are either automatically isolated, normally closed or contained within a tertiary containment.

PSC has concluded that the containment isolation provisions at Fort St. Vrain do not require modifications.

TABLE 2.1.4-1 CONTAINMENT PENETRATIONS AND ISOLATION PROVI5 IONS NUMBER SIZE- 10 CFR 50 CRIT ESSENTIAL ISOLATION ISOLATION SERVICE OF LIFES CLASS 55 56 57 YES NO AUTO MANUAL SIGNAL REMARKS SOURCE

1. Cenhol Rod Drive He 37 3/4-D32 x check Check valve integral part of PI-II-l

,{j) valve penetration ; remnte manaal isolation Purg? Lines valve outside containment

2. Reserve Shutdown Hopper PI-11-1 3/4-D5 x x(l) check Check valve integral part of Pressurization Lines 37 valve renetration. remote manual isolation valve outside containnent 6"-A3 x x(5) x Rupture disc and safety valve down- PI-il-5

3. PCRV Safety Valve Line 1 stream of isolation valve.

4. He Circulator Interspace PI-11-4 x x x PPS/HS Pressurization Lines 4 3/4-D32

5. Steam Generator Interspace PI-ll-4 x x x PPS/HS Pressurization Lines 12 3/4-D32

6. He Circulator Penetration Overpressure Protection Rupture Disc d?wnstream of isolation F1-11-4 16 1"-D-32 x x(5) Remote HS Lines valve

7. Stream Generator Penetration Overpressure Protection Rupture Disc downstream of isolation F1-11-4 Lines 12 4"-D3 x x(5) Remote HS valve

8. Core Support Floor Vent See A norwally closed pressure control P1-11-5 1"-A3 x x Renote HS Lines 12 Rema rks valve controls discharges to the gas water system

9. Core Support Floor See A normally closed pressure control valve 1"-A3 x x Remote HS Column Vent Lines 12 Roma rks controls discharges to the gas waste sys.

PI-ll-5

10. Plateout Probes Penetration Interspace x x Auto Isolation valve in supply Press. Lines 2 3/4"-D32 x x header isolates on high pressure or flow or manual remote PI-11-3

11. Process & Holsture Instrument Penet. " " " "

PI-II-3 Press. Lines 6 3/4"-D32 x x x x

TAPI E 2.1.4-1 (CONT 1f:UID)

I:lt EE R SIZE- 10 CfR 50 CPIT ESStNTIAL ISOLATION ISOLATION SE RVICE ef tlNFS (IASS 56 57 VES N0 f,U T O f t/ hu/L SION/L FEf7Fts SitTtE

12. Dutlet Coolant Therne- 7 3/A".D32 x x x x Same as 810 PI-II-3 couple renet. Press.

Lines

13. Circulator Instrupentation Penet. Press. Lines 2 3/4"-D32 x x x x PI-11-3

14. Thermocouple Penetration Press. Line 1 1/4"-032 x x x x " " "

PI-11-3

15. Spare Instrumentation Penet. Press. Line 2 3/4"-D32 x x x x " " "

PI-11-3

16. Bottom Access Penetration Press. Line 1 3/4"-C5 x x x x " " "

PI-11-3

17. Upper Access Penetration Press. Line 1 3/4"-D32 x x x x PI-11-1

18. Pressure Sensor Test Connections 6 ("-Bil x x Check x Normally closed .runual IC-II-2 Valve isolation valve

19. Sample Rake Test Conr.ection 4  %"-Bil x x Check x Normally closed manual IC-11-2 Valve isolation valve

20. Analytical Instrumentation 2 5"-B11 x x x RIS-7312 Closes on high radiation IC-11-2

21. Fast Gas Sample Lines 3  %"-Bil x x x RIS-7312 Closes on high radiation IC-11-7(1)

IC-11-2(2)

22. Holsture Monitor Sample 26  %"-Bil x x x Contained entirely within 10-11-2 Lines tertiary containnent of instrument penetration

23. Reactor Pressure to 1 5"-B11 x x x Norwully closed isolation 1C-11-2 Range Instr.(PDEll56) valves.

24. Core Pressure Drop 2  %"-B11 x x x Contained entirely within 10-11-2 Sensing Lines tertiary containment of instrument Penetration.

25. PCRV Safety Valve Tank 1 3/4"-A3 x x Remote HS Normally closed / fail PI-11-5 Vent Line closed isolation valve.

26. PCRV Safety Valve Isolation 2 2/4"-B4 x x x Packing gland provides PI-11-5 Valve Stem Packing Gland Vent secondary containment pressure boundary

27. PCRV Safety Valve Test 2 1"-A3 x x x locked Closed PI-11-5 Connection PI-11-5

28. PCRV Safety Valve Line 1 3/4"- A 3 x x x Excess flow check valve to Instrumentation closed instrumentation Check Check valve inside containment, PI-II-5

29. Purified He Supply to 1 1"-A3 x x PCRV Safety Viv. Line Valve rrnote manual isolation valve outside containnent

30. DELETED

TABLE 2.1.4-1 (CONTINUED) .

NUMBER SIZE- 10 CFR 50 CRIT ESSENTIAL ISOLATION ISOLATION SGHCE QF_ U NES CLA.SS 55 56 SZ__IES ____M_lul0 PiWJAL___ _ SIGNAL _ _REMAFIS _ _ _ __ _ d OURLE___ ___

31. Circulator inlet 8 3/4"-B5 x x x x Auto Isolation valve in supply PI-11-3 Thermocouple Well Press header isolates on high pressure Line or flow or manual remote

32. R/P Cooling Mater to

!!.T. Filter Adsorber 4 2"-D32 x x x locked clused PI-23-1

33. Purification Cooling 4 25-D13 x x(3) Remote HS PI-23-1 Water to He Purif.

Cooler

34. He P.,rification Cooler 2 3/4-B4 x x x Locked Closed PI-23-1 W il Vent Lines

35. He Purification Dryer 2 3/4-B4 x x EDSH- Closes en high f sow.Secendary PI-23-1 Drain Lines z 31< 3 isolation is nomally closed and interlocked.

36. He Dryer Regeneration 2 2"/15-A3 x x PDSH- Clases on high flow. Secondary II-23-I Return Line 23123 isolation is nomally closed and interlocked.

37. Purification Wells 6 1"-B4 x x x Normally clo ed PI-23-1 Vacuum Connections

38. Purification Train 2 1/8" x x x RIS-7312 Closes on high radiation PI-23-1 Activity Sampling Lines

39. LN2 Supply / Return to Low Temp. Adsorber 4 2"/1"-Dil x x(3) Remote HS PI-23-1

40. Regeneration Lines- 4 2"/15"Al x x PDSH- Closes on high flow. Secondary II-23-l Low Temp Adsorber 23123 isolatinn it nomally closed

41. Purified He 2 2"D32 x x(3) and interlocked. p g _2 3_ j x PDSH-23123 Closes on high flow

42. He Purification Cooler 6 3/8"B11 x x(3) x PDSH-Bubbler Lines Closes on high flow 23123

43. He Circ / Steam Gen. 6 3/4"D32 x x x locked Closed PI-ll-4 Interspace Drains

44. He Circ. Steam / Water 4 15-013 x(2) x PPS/HS PI-21-7 thrul0 Drain y

4 3/4"D32 x(2)

45. He Cire. He/ Water Drain Pressure Sensor x PPS/HS 4 1-1/3-D13 x(2)

46. He Circ. Water Drain x(2) Check Check valve inside containwnt

47. He Circ. Water feed 4 15-030 Valve remote / auto stop (heck valve outside x PPS/HS 15-A3 x(2)

48. He Circ. He/ Water Drain 4 4 5"D35 x(4) Remote Hs Normally closed and interloded

49. He Circ. Shutdown Seal Feed

TABLE 2.1.4-1 (CONTINUED)

NUT,ER SIZE- 10 CFR 50 CRIT ESSENTIAL ISOLATION ISOLATION SERVICE OF LINES CIASS 55 56 57 YES N9 AUTO P%NUAL SIGNAL REMARKS SOURCE

50. He Circ. Brake feed 4 '"D35

. x(4) Remote 11 5 Norinally closed and Pl21-7 thru 10 interlocked

51. ';e Circ. llellum 4 l '," A 3 x(2) Check Check valve inside containment Buffer feed Valve remote f auto isolation valve outsfo.

52. He Circ. He Buf fer 4 3/4*D32 x(2) x Pressure Sensor

53. He Circ. Water reed 4 1'2"D30 x(2) Check Check valve inside containment Valve manual isolation valve outside

54. lie Circ. Labyrinth 4 3/4"D32 x(2) x Pressure S- ,-

55. He Circ. h, Water Drain 4 l'3"A 3 x(2) x PPS/HS

56. He Circ. Feedwater 4 3/4*D13 x(2) x Pressure Sensor

57. lie Circ. High Pressure 4 l'3"D13 x(2) x PPS/HS Water Drain

58. He Circ. Steam / Water 4 3/4"D13 x(2) x Pressure Sensor

59. He Circ. Turtine Exh. 4 3/4"D9 x(2) x Pressure Sensor 60 He Circ. Drain 4 3/4"D13 x Capped Pressure Scnsor

61. ' e Circ. Guard lie Feed 4 1"AJ x(2) Check Check valve inside containment Valve remote / auto isolation valve outside

62. He Circ. Water Turbine 4 6"D28 x(2) x PPS/liS Feedvater Supply

63. lie Circ Water Turbine 4 10"D13 x(2) x PPS/HS Feedwater Drain

64. Steam Generator feed- 12 4"D28 x x(2) Remote 11 5 PI-22-1,6 Water PI-22-2,7 S. Steam Generator Super- 12 83/4/14"D1 x x(2) x FPS /its heater Outlet PI-22-3,8

66. Steam Generator 12 16"/12"D9 x x(2) x PPS/HS Cold Reheat PI-22-4,9

67. Steam Gererator Hot 12 ll"/22'."D3 x x(2) x PPS/HS Reheat

68. Circulator LP and Flow 16 '."-B i l x x x Contained enti mly with- 10-11-3 Instrumentation in tertiary containment of instrument penetration.

TABLE 2.1.4-1 (CONTINUED)

NUMBER SIZE- 10 Crr 50 CRIT ESSENTIAL ISOLATION ISOLATION SERVICE OF LINES CLASS 55 56 57 YES NO AUTO MANt1AL SIGNAL REMAR_KS ___ _SOU RCL _ _ _ ___

69. High Temp. Filter 2 3/4"B5 x x x x Auto isolation valve in PI-11-1 Adsorber Penetration supply header isolates on Purge high pressure or flow or manual remote.

70. High Temp. Filter 2 3/4"D32 x x Check Check valve outside containment PI-11-1 Adsorter Penetration valve Various small instrument lines Purge terminate at pressure sensors.

All lines have manual isolation v 31 ves .

71. PCRV Liner Cooling 20 4"/6"C3 x x(2l x High Pressure PI-46-2 or low flow scanner (1) Essential for Reactivity Control (2) Essential for Core Cooling (3) Essential for Depressurization (4) Essential for Containwent (5) Essential for Overpressure Protection

Section 2.1.8.a Improved Post-Accident Samplino Capability The following question was raised by the Ti1I-2 review team during the January 21-22, 1980 site visit:

1. Provide a description of modifications to the sampling system (if none, say so) and a description of the new radiochemical analysis facility.

PSC Reply:

No modifications have been made to the FSV sampling systems.

The new radiochemistry laboratory is to be relocated to a concrete-shielded building to be constructed adjacent to the Reactor Building.

The new radiochemistry laboratory will contain all the equipment presently installed in the existing laboratory and the analytical capabilities will be the same.

The new radiochemistry laboratory will have appropriate ventilation and waste processing capabilities.

Section 2.1.8.b Effluent bbnitors The following questions were raised by the Tl11-2 review team during the January 21-22, 1980 site visit.

1. Provide information specified in Denton's October 30, 1979, letter regarding both noble gas and iodine / particulate monitoring.

2. Implement procedures for noble gas monitoring (interim).

3. State that procedures for estimating particulate releases have been implemented (in writing).

PSC Reply

1. In our December 28, 1979 response (P-79312) we addressed the subjects of noble gas and iodine monitoring with the conclusion that only one monitor (RT-93251-1) required upgrading. We committed to change out this monitor by January 1,1981. In our October 29, 1979 response (P-79249) we provided as an attachment a copy of our letter to the NRC dated June 15,1979 (P-79130) which presented in tabular form a description of the effluent monitors. This tabulation, along with our supplementary responses referenced above, in our opinion addresses items (1), (ii), (iii) and (iv) on pages 33 and 34 of the NRC October 30, 1979 letter. For item (v) all of the monitors listed in the tabulation of P-79130 are powered off an essential bus except RT-4803 (Ping tionitor). RT-4803 is on the ACM bus, however, which could serve as an essential power supply should the ACM diesel generators be put in service. The Ping !!onitor (RT-4803) is useful in accident situations, but is not essential. All monitors listed in P-79130 provide direct control room readout, again with the exception of the Ping Monitor (RT-4803).

Although we have indicated in our December 28, 1979 response that our noble gas and iodine monitors will remain on scale during accident conditions, PSC has installed, per our December 12, 1979 response (P-79299), temporary portable high range monitoring equipment.

This instrumentation consists of a Cutie-Pie Ptodel 2588 with a flodel 2520 probe with a range of 0 - 2.5 R/hr, and an Eberline flodel E-5008 with a modified liodel HP-177 probe providing a range of 0 - 200 mRr/hr. These instruments are calibrated on a six (6) month frequency utilizing established procedures. Procedures have been developed and are in place to provide an estimate of release rates based on dose rate readings from the instruments listed above. Energy Response Curves for these instruments are attached.

Although PSC has indicated that the temporary high range monitors are not required at Fort St. Vrain, PSC intends to leave these instruments in place until we receive NRC concurrence with the position we have taken regarding our permanent monitors.

2. The procedures for the temporary high range noble gas monitors have been issued.

3. Procedures for estimating particulate releases have been issued.

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/j po 4 " (U931Lm30u.hG public service company ce ccEondo

?. O. Box 361, Platteville, Co 80551 June 15, 1979 Fort St. Vr'd-Uni No. 1

19130 F.r. George Ku..=yce, ?roj ect F.anager U.S. Nuclear Regula

ory Ce dssion Division Projec: F.anage=en Special Projects

'4ashington, D.C. 20535

Subject:

Gaseous P.ffluen: F.cniters

Dear F.r. Ku==yc:

Per your infor=ed inqui-y concerning the radia: ion =eni: ors at Fort St. Vrain, we have prepared the attached :able in response to the eight (8) questicus you forwarded.

'Je have taken a quick look at the high-range stack =c=1:cr, RI-7324-2, and have concluded that if we were :o release the entire pr -ary coolant inventory ou: the plant stack ass 'mg a 30,000 C1 d

design circulating, activity, our high-range stack =ent:c: vould re-

=ain on scale (N10" cpm versus a capabili:7 of 107 eps for the =cnitor) .

If you need any fur.her infor=atics, please let =e knew.

Very truly yours, WW!'

Don '4. 'Jare= bourg Id.anager, Nuclear ?redue:1cn D'n~4/vt Esc.

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tation panel.

4-Cit 5-Turbine Deck Areu

• Yl.eue calculations are based on a reactor building exhaust flow of 1.6E7 cc/sec.

• • Calibration Factors revised slightly from June 15, 1979 submittal (P-79130) as a result of reevaluation.

Section 2.1.8.c Portable Iodine ibnitoring The following question was raised by the TMI-2 review team during the January 21-22, 1980 site visit:

1. Implement the requirements of this item (10 minute turn around time)

PSC Reply:

The Tt11-2 review team questioned our capability for analyzing samples with a ten (10) minute turn around time. This time appea: s to be a new requirement which is not included in any of the NRC written requirements.

As indicated in our December 28, 1979 response (P-79312) we have developed procedures for taking air samples utilizing charcoal adsorbers and have relocated one of two multi-channel analyzers to a low background area to permit counting of these charcoal filters. Utilizing the most restrictive Iodine MPC, we have demonstrated the capability for a ten (10) minute turn around time for analysis after the sample is taken.

We have identified the areas that require continuous habitability as the Control Room, the Technical Support Center /0n-Site Command Post and the Operational Support Center / Personnel Control Center. We have also identified the need to take samples in the reactor building should access to the reactor building be required. We are presently incorporating these areas into procedures which define the various action levels for taking samples and the frequency for taking samples. These procedures will be in place prior to the resumption of reactor power operation.

Section 2.2.1.a Shift Supervisor's Responsibilities The following questions were raised by the Tl1I-2 review team during the January 21-22, 1980 site visit:

1. Provide written direction to the Shift Supervisor not to be involved in individual operations at the expense of overall plant operations.

2. Clearly specify, in writing, who (by title or qualification) can relieve the Shift Supervisor during accident conditions.

3. Document the review of Shift St isor responsibilities and verify that it will occur periot.: . ally.

PSC Reply:

1. Specific written direction has been issued. See attached Production Department Policy, which has been issued in the form of an Operation Order.

2. An Operation Order that incorporates the policy statement has been issued which clarifies the individuals that can relieve the Shift Supervisor during accident conditions (see Producticn Department Policy, attached). This same Operation Order which incorporates the Production Department Policy letter has been revised to address the chain of command in the control room and temporary bsences of the Shift Supervisor.

3. A surveillance requirement (SR) is being established to ensure an annual review of Shift Supervisor responsibilities by Plant Management and the Vice President of Production.

This SR will be controlled and scheduled similar to Technical Specification SR's to ensure an annual review and documentation of this review. This SR procedure will be implemented prior to resumption of reactor power operation.

. ATTACHMENT Section 2.2.1.a N

\

PUBLIC SERVICE COMPANY OF COLORADO P. O. 80x 840 DENVER, COLORADO 802o8 OSCAR R. LEE January 1, 1980 vice ncsioc~r Revised - February 5, 1980 ATTACHMENT II PRODUCTION DEPARTMENT POLICY Shift Supervisor's Duties Fort St. Vrain Station The prime responsibility of the Shift Supervisor on duty at the Fort St. Vrain Station is to manage the operation of the plant in such a manner that it is always in a safe operating condition.

He has the authority to maintain a safe operating condition, without regard to system demand or requests from load dispatcher (or others) for generation.

This responsibility, and corresponding authority, takes precedence over those duties outlined in the job description for Shif t Super-visors, and must be always the first consideration when determining operating conditions.

Under radiological emergency conditions, the Shif t Supervisor is assigned the duties of Emergency Director. In this position, the Shift Super-visor shall report to the control room and remain in the control room until properly relieved by another qualified SRO. Except for unusual circumstances beyond normal administrative control the Shif t Supervisor may be relieved only by another Shift Supervisor or by the Superinten-dent of Operations.

In the position of Emergency Director, the Shift Supervisor assumes the duties of the Emergency Director as specified in the Station Emergency Plan. The primary duty of the Shift Supervisor under emer-gency conditions is overall plant operation and emergency coordin

• ion.

The Shift Supervisor shall not become involved in individual operat. 3nal problems or activities at the expense of overall plant operations.

<5/$uu

0. R. Lee ORL:dkm

Section 2.2.1.c Shift Turnover Procedures The following questions were raised by the THI-2 review team during the January 21-22, 1980 site visit:

1. Verify that valve positions, indicated in the control roum, for essential systems not in use are checked for proper position during shift turnover.

2. State method, including enumeration of procedures, of implementing the requirements of part 3 of Section 2.2.1.c. Include assurance that this will be checked annually.

PSC Reply:

1. An investigation of the essential plant system valves that have position indication facilities in the Control Room has been conducted. This investigation indicates that the Fort St. Vrain facility does not have any essential valves, which are not normally in use, that would fit the Till-2 review teams criteria for valves that require a separate check system for

" shift turnover" use.

In an effort to provide better continuous monitoring of the conditions of all facility systems, we have implemented some changes to existing procedural methods used to provide system monitoring. These changes include the addition of a " System Abnormality Log" which will include, by system, all " clearances",

and all " system deviations". This log book will be maintained by the Reactor Operators for the purpose of providing ready access to all information concerning abnormal plant conditions.

The review of this log shall be a required part of the " shift turnover" process. Information contained in the " System Abnormality Log" shall be duplicated by the use of multi-part forms for use by the Shift Supervisor and available to the Shift Supervisor in his office. The System Abnormality Log will be in place prior to resumption of power operation.

2. The shift relief turnover procedures are included in the QA surveillance and audit program and additionally in the NFSC audit program. The operational QA program provides routine QA surveillance of all edministrative controls and specifically under QA Surveillanct Procedure, QASP-301, Plant Operations, the shift turnover procedures are included. Logs, opertional records, and specific valve positions are checked on a random basis. Portions of this surveillance program are conducted on at least a quarterly basis. The QA audit program provides additional checks in this same area. The QA audit program is scheduled to provide complete coverage of all aspects of the QA program once every two years.

The combination of the QA surveillance program, the QA audit program and the NFSC audit program provide independent checks of plant operations and specifically shift turnover procedures on a more than adequate frequency to meet the annual requirements of the October 30, 1979 NRC letter.

Section 2.2.2.b Technical Support Center /0n-Site Command Post In our October 29, 1979 response (P-79249), the interim Technical Support Center /On-Site Command Post (TSC/0CP) was described. On January 21 and 22 the Ti1I-2 review team requested that we address the following:

1. Provide an update description of the TSC/0CP (Reference to NRC Hotline)

2. Provide dedicated communications between the TSC/0CP and the Control Room

3. Provide Technical Data in the TSC/0CP

4. Verify that best efforts have been made to directly display parameters in the TSC/0CP

5. Implement requirements of 2.1.8c including procedure modification PSC Reply:

1. The attached Figures 1 and 2 provide the location of the interim TSC/0CP with reference to the Control Room and the Shift Supervisor's office. In our October 29 response we indicated that the interim TSC/0CP would be equipped with commercial telephones, back-up radio communications, and necessary technical data (i.e., P & I diagrams, Electrical Schematics, FSAR, Technical Specifications, and Emergency Procedures).

With reference to communications in the TSC/CCP, the interim TSC/0CP is presently equipped with three (3) commercial telephones (each phone is a five (5) button set with five (5) phone lines), One (1) NRC Hotline, one H.P. Hotline (the NRC Hotline and the H.P. Hotline were installed by the NRC) and provisions exist for back-up radio communications (battery operated) between the TSC/0CP and the Control Room.

Overall communications for the various emergency centers are shown on Figures 3 and 4.

2. With reference to dedicated communications between the TSC/0CP and the Control Room it was our position that due to the close proximity of the TSC/0CP to the Control Room that personnel liaison would be utilized as a primary means of communications.

Emergency Procedures were developed to establish personnel liaison utilizing the second Technical Advisor (TA) as liaison between the TSC/0CP and the Control Room. This was established and is consistent with the NRC position that the Shift Supervisor should not be tied up with specific operation and has the primary responsibility of overall plant operations. Personnel liaison between the Technical Advisor in the interim TSC/0CP and the Technical Advisor in the Control Room or the Superintendent of Operations in the Control Room avoids distractions of the Shift Supervisor and certainly provides for adequate communications.

Back-up communications are also availtbl. via the telephones or the radio. We feel that this commt nication system is more than adequate to meet the intent of the TSC/0CP Communications requirements, however, because we seem to be at an impasse, we

are installing an interccm system between the interim TSC/0CP and the Control Room to serve as additional communications to meet the intent of the October 30 letter. This intercom system will be installed prior to resumption of reactor power operations at Fort St. Vrain.

3. All technical data specified in our October 29, 1979 response (P-79249) has been placed in the TSC/0CP. This data includes P & I Diagrams, one-line electrical schematics, FSAR, Technical Specifications, Emergency Procedures, Standard Operating Procedures, Operations in Degraded Mode Procedures and Fire Control Procedures.

4. Again, due to the close proximity of the interim TSC/0CP to the Control Room, we have not provided any direct read-out parameters in the interim TSC/0CP. We have estab'ished access to the Control Room by TSC/0CP personnel and liason is established by the Technical Advisors. Critical plant parameters are readily accessible.

5. Concerning radiation monitoring, the interim TSC/0CP is adjacent to the Control Room and the TSC/0CP atmosphere is controlled by the same ventilation system and the same shielding as the Control Room. There is an area radiation monitor in the Control Room and a continuous air monitor in the Control Room ventilation system both of which alarm / read-out in the Control Room. Alarms or activity experienced with these monitors shall be utilized as initiating action to ta}gyfurther air samples in the TSC/0CP and Control Room for I. Your October 30 letter indicated that protective measures should be described if levels become too high such as to warrant evacuation of the TSC/0CP to the Control Room. In our case the TSC/0CP is essentially part of the Control Room. Again referencing our earlier response, we do not have an iodine problem at Fort St.

Vrain. As indicated in Section 14.11 of the FSAR, DBA-2, even for maximum depressurization of the primary coolant the expected Iodine release is 10 curies with a design inventory of only 60.8 curies. Loss of Forced Coolant, DBA-1, does not result in any significant Iodine releases to the reactor building.

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Section 2.2.2.c Operational Support Center During the recent visit of the THI-2 review team, we were requested to provide additional information concerning the Operational Support Center as follows:

1. Implement requirements of paragraph 2.1.8.c of the October 30, 1979 NRC letter.

PSC Reply:

As indicated in our December 12, 1979 response (P-79299), we propose to check areas which require continuous habitability during accident conditions with air samplers utilizing charcoal filter adsorbers. Samples are to be analyzed utilizing a multi-channel analyzer.

Areas which require continuous personnel habitability are defined as the Technical Support Center /0n-Site Command Post, the Control Room, and the Operational Support Center / Personnel Control Center.

As indicated in the response to Section 2.2.2.b, we will utilize the radiation monitors in the control room ventilation system to trigger action for taking iodine samples in the Control Room / Technical Support Center /0n-Site Command Post.

For the Operational Support Center / Personnel Control Center which is outside the plant, we will utilize indication from our fodine stack monitor to trigger the need for iodine sampling in the Operational Support Center / Personnel Control Center.

Procedures which implement the above requirements will be in effect prior to resumption of reactor power operation.

As a point of clarification, we indicated in our December 12, 1979 response that plant personnel report to their specific emergency stations as a first action. The MPC levels for unidentified nuclides in plant, however, are sufficiently low to ensure that personnel will be evacuated from the various emergency stations long before iodine concentrations become a problem.