ML20215A752

SER Supporting Util 860523 Proposed Changes to Tech Specs Re Reactor Trip Sys Instrumentation & Surveillance in Response to Generic Ltr 85-09.SALP Input Also Encl

ML20215A752
Person / Time
Site:	Sequoyah, Vogtle, 05000000
Issue date:	07/07/1986
From:	NRC
To:
Shared Package
ML20213E629	List: ML20213E625 ML20213F117 ML20214G397 ML20214H017 ML20214H085 ML20214H125 ML20215A123 ML20215A344 ML20215A593 ML20215A752 ML20215A782 ML20215A967
References
GL-85-09, GL-85-9, NUDOCS 8612110386
Download: ML20215A752 (33)

Text

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SAFETY EVALUATION REPORT

_ TECHNICAL SPECIFICATION CHANGES (MPA B-90)

SEQUOYAH NUCLEAR STATION, UNITS I AND 2 INTRODUCTION By letter dated May 23, 1986, TVA proposed changer to the Sequoyah Technical SpecificationspertainingtoReactorTripSystemInhtrumentationandSurveillance in response to Generic Letter 85-09 (MPA B-90). Generic Letter 85-09 concluded that Technical Specification changes should be proposed by licensees to explicitly require independent testing of the undervoltage and shunt trip attachments of the reactor trip breakers during power operation, testing of bypass breakers prior to use, and independent testing of the control room manual switch contacts and wiring during each refueling outage.

EVALUATION TVA's proposed Technical Specification changes are consistent with those specified in Generic Letter 85-09, except as indicated below by items 1,2 and 3.

1.

A different Action than that indicated is needed for Modes 3*,4* and 5*

(asterisk means with the reactor trip system breakers in the closed posi-tion, the controi rod drive system capable of rod withdrawal, and fuel in the reactor vessel) for Functional Units 20 and 21 of Table 3.3-1.

2.

No provision is included for testing of the Reactor Trip Bypass Breaker prior to each time it is put into service.

3.

No provision is included for verifying the operability of the Bypass Breaker trip circuits each refueling outage.

8612110386 861205 PDR ADOCK 05000424 A PDR

The proposed Technical Specification changes for Functional Units 20 (Reactor Trip Breakers) and 21 (Automatic Trip Logic) apply the same Action 12 to all Applicable Modes. Action 12, which requires going to Hot Standby under certain conditions, is not appropriate for Modes 3*,4* and 5*.

For Modes 3*,4* and 5*,

tha-reactor trip breakers should be opened if the condition is not remedied within a given time period as specified in Action 13 of Generic Letter 85-09.

TVA has declined to provide Technical Specification changes for surveillance testing of the reactor trip bypass breakers on the basis that Westinghouse Owners Group (WOG) calculations have shown no significant reliability improvement from including periodic surveillance tests of the bypass breakers in the technical specifications. This is not acceptable to the NRC staff. When a reactor trip breaker is tested, the reactor trip bypass breaker is put into service in its place. The Generic Letter 85-09 Technical Specifications specify (Table Notation 13) a manual test of the bypass breaker (either a local shunt trip or remote undervoltage trip) prict to putting it into service. This testing is a simple procedure and it is prudent to do this test before relying on the breaker.

The Generic Letter 85-09 Technical Specifications (Table Notation 11) also re-quires a manual trip function test to independently verify the Operability of the undervoltage and shunt trip circuits of the Bypass Breakers each refueling outage. This test is important to assure that the bypass breaker' can be manu-ally tripped from the control panel.

I

t CONCLUSION '

The licensee should amend its proposed Technical Specification changes by

1. proposing a different Action than that indicated for Modes 3*,4* and 5* for Functional Units 20 and 21 of Table 3.3-1,

2. including provisions for testing the Reactor Trip Bypass Breaker prior to each time it is put into service, and

3. including a provision for verifying the operability of the Bypass Breaker trip circuits each refueling outage.

Each of the above changes should be in accordance with Generic Letter 85-09.

O G e

i

EICSB SALP INPUT i .

PLANT: S quoyah. Units I and 2

SUBJECT:

Proposed, Technical Specification Changes, MPA B-90 i

EVALUATION PERFORW.NCE 1

CRITERIA CATEGORY BASIS

. Management N/A. ggo- basis for assessment.

~

Involvement

. App M to 2 A1,gough not go,nsistent with staff position, supporting justification I Resolution of. .

was prov1.ded, -

! Technical Issues 1

Responsiveness 2 i .

,gn_se was, adequately supported and consistent with time schedule.

i 1

. Enforcement History s N/A

No basis for assessment.

i, -

[ .

R* Portable Events , N/A No basis for assessment. -

l . Staffing N/A

! No basis for assessment.

l ^

l

. Training N/A No basis for assessment.

e j# \

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_ _ . I

Issue (b)

Leak Testing of Pressure Isolation Valves Request for Additional Information 210. Mechanical Engineering Branch f.

The Surveillance Requirement pertaining to leak testing of pressure isolation valves (PIVs) presented in Section 4.4.3.2.2 of Perry Draft Technical Specification is not complete. In addition to the two requirements currently identified in Perry draft Technical Specification, Section 4.4.3.2.2, the staff requires the PIVS to be leak tested (a) prior to entering the Hot Shutdown whenever the plant has been in Cold Shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more and if leakage testing has not been performed in the previous 9 months and (b) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following valve actuation due to automatic or manual action or flow through the valve. Provide additional information to assure that the Perry plant has the following plant features: (1) full closure of PIV's is verified in the control room by direct monitoring position indicators, (2) inadvertent opening of PIV's is prevented by interlocks which require the primary system pressure to be below subsystem design pressure prior to openings, and (3) gross intersystem leakages into the low-pressure core spray, residual heat removal / low-pressure coolant injection, and residual heat removal / shutdown cooling return and suction lines would be detected by high-pressure alarms and increases in the suppression pool level.

With these plant features in place, the PIV's are controlled and verified continuot. sly rather than at the intervals specified in (a) and (b) above and then, the exception for relief from the surveillance requirements (a) and (b) could be accepted. "

t The following is a list of sections of the proposed Vogtle Technical Specifications'that received comments. If a section did not warrant any comments, it will not be noted below.

Technical Specification Comment Index Page IV: Table 3.1-1 Title does not agree with actual title on. table.

" FULL-LENGTH ROD should be "CO. NTROL.OR SHUTDOWN ROD" Index Page V: Typo " Table 3-3-4" should be

" Table 3.3-4" Index Page VI: The title " REMOTE SHUTDOWN INSTRUMENTATION" should be " REMOTE SHUTDOWN SYSTEM" to agree with title on 3/4 3-47 or vice versa.

Index Page VI: The title for page 3.3-7 does not agree with the title on page 3/4 3-48. Comment related to the one above.

Index Page VI: No Tech Spec for Loose Part Detection System.

Index Page VII: Typo "safetry" should be

" safety" Index Page X: Title of TS 3/4.7.6 does not agree with the title of the actual TS.

l Index Page X: Figure 4.7-1 title needs to be clarified "2)"

Index Page XIV: Table 4.12-1 needs complete title.

Index Page XVIII: Section 3/4.7.7 title does not agree with title on page B 3/4 7-4.

Index Page XIX: Section 3/4.9.6 title does not agree with title on page B 3/4 9-2.

Index Page XIX: Section 3/4.9.12 listed does not have a basis in TS.

3.1.3.2 This TS as written only applies to the control rod positions. This should be shutdown and control rod positions for clarification as done in TS 3.1.3.3.

1

3.1.3.3 The

• indicating "with the Reactor Trip '

, System breakers in the closed position" appears unnecessary in that the action statement would control with the standard phrase The provisions of 3.0.4 and 4.0.4 would apply.

Table 2.2-1 A number of the values given here are discrepant with a letter dated September 29, 1986, f' rom the Vogtle Project to Mr.

Denton entitled WCAP 11269 (proprietary) and 11270 (non proprietary),

" Westinghouse Setpoint Methodology for

- Protection Systems - Vogtle Station" Since this letter is addressed to NRR, we will not compare them directly except in areas that we feel may still be unclear. One such example concerns Functional Unit 16, Turbine Trip. The WCAP report leaves theso values blank while the proposed TS value for Turbine Stop Valve Closure is extremely non-conservative with respect to STS (DRAFT) Rev. 5 which lists >1% open for both the Trip Setpoint and Allowable Value while the proposed Vogtle TS lists

>97.6%. Additionally, Notes 1 & 3 of the

, proposed Vogtle TS and the WCAP values denote variable time constants while the STS uses equal signs. Also, the

• symbol means two different items.

/

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3.2.2 Remove brackets around constants in i equation as these are site specific l numbers.

l 3.3.1 & 4.3.1.2 The surveillance calls for checking the response times within the limits while the limits are no longer a part of TS 3.3.1. If these limits are to be in the FSAR, then the proper references should be made.

3.3.2 & 4.3.2.2 Same comments as above regarding l response times.

Pressurizer pressure-low, reactor

! coolant pump-under voltage, and reactor l coolant pump underfrequency are only applicable above the P-7 interlock. This I is not as conservative as the guidance l

2

l

'_ h Table 3.3-1 (cont'd) given in STS. Action statement 2a. uses .

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> instead of the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> used in .

- STS. Action statement 2b. uses 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> instead of the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> used in the STS.

The same applies to 6a. and 6b. Action statement 12 also states 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> instead of the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> used in the STS. Action statement 13 is non-conservative and has ATWS implications.

Table 4.3-1 It should be noted that there is no table given for Reactor Trip System response times. Nearly all of the functional units in the Reactor Trip

- System Instrumentation Surveillance Requirements replaced the monthly surveillances for the analog channel operational test given in the STS with quarterly or refueling or startup surveillances. This is a matter that should require some sort of justification for making such a sweeping non-conservative alteration from STS.

Table 3.3-2 The ESFAS instrumentation table has no functional unit heading or channel trip criteria for Steam Line Flow High with or without Lo-Lo T-ave. Neither is there a functional unit for Under Voltage to the RCP's starting the Turbine Driven AFW pumps. This whole table is weak with respect to the AFW system. Additionally, there is no functional unit dealing with an ESF start due to steam generator water level or Lo-Lo T-ave.

Table 3.3-3 The values in this table were also addressed in the WCAP 11269 report transmitted September 29,1986 to B.J.

Youngblood from the Vogtle Project.

There were no problems encountered with this WCAP data.

Table 4.3-2 It should be noted that there is no table given for the ESF system response time. There are similar discrepancies in i this table as there were in Table 3.3-2.  :

There are no funtional units covering the surveillance of differential pressure between two steam lines, high steam flow in two lines coincident with Lo-Lo T-ave, no undervoltage to the RCP's, nor any ESFAS interlocks at P-12 i or P-14.

3

Table 3.3-8 This table typically addresses,' but does not include: .

-PORV position indication "

-PORV Block position indication

-Safety Valve Position

-Plant Vent Monitors

-Containment Isolation Valve Position Table 3.3-10 There is no.9riteria given for other process monitors'such as: Main steam reliefs and atmospheric discharge; Auxiliary feed pump turbine exhaust; ,

Containment purge system; Turbine

- building ventilation exhaust; Fuel storage area. Do these monitors exist?

3.3.3.8 The loose part detection TS is missing.

Either add or renumber remaining TS.

This should be added back.

4.4.1.2.2 , These TS all reference secondary side-4.4.1.3.2, & water level values at 17% which is 3.4.1.4.1b non-conservative per WCAP 11269. See general comments.

3.4.1.4.2 Typo "RHR" has been deleted before the word " trains" in action statement a.

3.4.2.1 Clarification- TS 3.4.1.4.2 changed the terminology from "RHR Loop" to "RHR Train". This TS needs the action statement changed to reflect consistent terminology.

4.4.3 The surveillance should add the following step: The emergency power supply for the pressurizer heaters should be demonstrated operable at least once per 18 months by manually transferring power from the normal to the emergency power supply and energizing the heaters.

i 4 i

+

3.4.4a The phrase,"because of e$cessive seat leakage" should be deleted as it will cause problems with the interpretation.

It should also be proscribed that the unit be in COLD shutdown, not HOT, within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, not 6. 3.4.4b

~

1 and 2 could be totally deleted with the above change.

3.4.4c Should.be revised to read,"With one or

~

more block valve (s) inoperable, within one (1) hour, restore the block valve (s) to operable status or close the block valve (s) and remove power from the block

. valve (s). Otherwise, be in at least HOT standby within the next six (6) hours and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. .

4.4.4.1 Should read; "In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated operable at least once per 18 months by:"

4.4.4.3 This specification should be added and should read,"The emergency power supply for the PORV's and block valves shall be demonstrated operable at least-once per 18 months by:

a) manually transferring motive and

. control power from the normal to the emergency power supply, and b) operating the valves through a complete cycle of full travel.

4.4.5.0 Should add;"and the requirements of specification 4.0.5 4.4.5.5.c Should read;" Results of steam generator tube inspections which fall into category C-3 and require and require .

Prompt notification of the commission ,

shall be reported pursuant to Specification 6.9.1 prior to resunption of plant operation. The written followup of this report shall provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence.

For C-3 results, " Action Required",

second paragraph should read," Prompt 5

i i

Table 4.4-2 (cont'd) notification to the NRC p,ursuant to specification 6.9.l'. Also.under 2nd sample inspection if Additional S/G is "C-3", action required shoul'd read," Inspect all tubes in each S/G and plus defective tubes. Prompt notification to NRC pursuant to

, Specification 6.9.1."

Table 4.3-6 Notes (4) E (5)'are acceptable IF manufacturers recommendations include

[)b ' using a one percent standard gas or equivalent thereof. (STS say use both 1%

and 4% for both Hydrogen and Oxygen.

Table 3.4-2 The limits in this table are too lenient for corrosion control and all PWR

, licensee's base their chemistry control on the much tighter limits in the

() guidelines recommended by the Steam Generators Owners Group /EPRI.

N 75re L

-Pable 3. 4-4 Table 3.4-4 is non-conservative with f)'d 4 respect to the values given in the STS.

. 4.4.9.1 There are no surveillance requirements on Reactor Pressure Vessel irradiated samples.

4.4.9.1.2 Specification deleted, why? ,

3.4.9.2c The maximum auxiliary spray water temperature differential is given as 625 ,

i degrees which is nearly twice the 320 '

degrees in the STS.

3.4.9.3 SAFETY CONCERN- Typically, reliance on the RHR relief valves is not appropriate since the suction valves will isolate to protect RHR at the sacrifice of the RCS.

The vent path option capable of relieving 670 GPM appears inadequate in that these TS no longer will require one of two charging pumps inoperable thus allowing for a potential of 1,000 GPM. '

It should also be noted that the vent path only has to be " capable of relieving" where the STS would require the path to be open and thus not require human intervention.

6

,-.----e

/l y

I k, '

.h-3.4.11 The site specific design has three i' <

valves per path, not two, that must be- ,

operable and closed to have a success' -

path, there are four isolation valves and two control valves. See FSAR figure .

5.1.2-2. . b ,

7e' r

4.4.11.1 There is no quarterly surveillance on _ Jy the operation of the RCS vent path. block valve. ,

'ki 3.5.3 Delete reference to instrument span in ) s1-3.5.1b. This will eliminate the i.

confusion between actual volume l requirements and what an instrument "g ',

reads.

\

4.5.1.1 STS has a provision for verifying every 18 months that the auto-open'on SI and ,

P-11 signals work as designed. This should be reinstated to be available in the event that the plant is in an action statement, or in 3.0.3 when an event 1 occurs. Why this was granted to Callaway '

is unknown, but does not justify carte '

blanche for Vogtle. >

I 4.5.2 LPSI system is not the appropriate title. Page 3/4 5-6 "LPSI" could be "CVCS" system. i l

l l 4.6.1.2 Reference to ANSI N45.4-[1972] most likely should be ANSI /ANS 56.8-1981. See FSAR section 6.2.6.

3.6.1.7.b Delete

• and note. This is part of STS guidance and not intended for the specification once the appropriate valve sizes are specified and hours are selected. If left as is, this implies '

that multiple paths can exist at one time. This simply is not the case.

7

5 4.6.1.7.2 See wr[ite-up above. If

• note remains, then each time "14-inch" appears, an *

- should be placed.

3.6.2.1 & 3.6.2.3 SAFETY CONCERN- The STS was written for a plant where the Containment Spray Systems and Containment Cooling Systems are redundant to each other. At Vogtle this is not the case. With one spray system out, you must have both cooling trains and vice versa. With this in mind, TS 3.6.2.3 Actions a,b.and c rely on the other system to provide the necessary cooling capability. It is suggested that a format such as TS 3.5.2 be utilized to combine TS's 3.6.2.1 & 3.6.2.3 and possibly 3.6.2.2 to reflect a cooling train concept and eliminate the excessive outage times.

I 3.6.2.3 1) See above. .

2) The applicant has only provided for 100% cooling capacity as designed, not as would be required to cool containment; the spray system is necessary to reach a 100% cooling capacity.

3) Delete," equivalent to 100% cooling capacity" This should be a guidance statement for developing TS and is not a Part of the LCO.

3.6.3 The isolation valves that this TS is referring to need to be referenced in TS i or bases.

3.6.6 1) renumber to 3.6.5

2) Applicant should provide data and not delete this specification.

3.7.1.1 1) 33/4.7.1 " Turbine Cycle *# , need to delete the extra 3.

2) The action statement forces plant to be in COLD SHUTDOWN vice HOT SHUTDOWN.

Either this is wrong or applicability should include MODE 4.

I Table 3.7.1 Under percent of rated thermal power, 65 should be 64 and 43 should be 42.

8

Table 4.7.1 SIS do not have a footnote. The sample and analysis frequency do not specify radioiodines.

3.7.1.1 Why not delete these tables and reference the appropriate sections of the FSAR as done elsewhere or at least combine to one page.

4.7.1.2.la. "on a staggerred test basis" should be l deleted. l 4.7.1.2.la.4) Should read," Verifying that each automatic in the flow path is in the fully open position whenever the Auxiliary Feedwater System is placed in automatic control or when above 10%

rated power.

3.7.1.3 Action a. should read," Demonstrate the operability of CST V4002 as a backup supply to the auxiliary feedwater pumps and restore CST V4001 to.....Also, delete the word vessel from the LCO.

3.7.1.5 & 4.7.1.5 A normal Westinghouse PWR has a check valve and one MSIV per steam line. The STS allows inoperability of the MSIV for four hours and relies on the check valve. This TS's action a allows both MSIV's (no check valve exists) to be inoperable for four hours and then only requires one to be restored. Both actions a and b appear innappropriate in a that they take credit for the added MSIV but do not reflect the deletion of the check valve.

This specification and its associated surveillance do not resemble STS at all and appear to be less conservative.

This whole Specification should be examined.

4.7.3 There is no surveillance requirement to verify every 18 months during shutdown, that each automatic valve servicing safety-related equipment actuates to its correct position on a Safety Injection test signal.

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~ . . - .. - --

9 3.7.5 .

The Action statement was designed for TS 3.7.5.a only. TS 3.7.5.b & c need action statements similar to that of 3.7.4 to prevent unnecessary shutdowns of the plant.

4.7.5.b Delete the words " required number of" from surveillance.

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4.7.6.a Remove [ ] around 80 (degrees).

4 4.7.6c,e,f,g Does testing in ANSI N510-1980 meet 4.7.7b,c,d,e,f requirements of Reg Guide 1. 5.2, Rev- 2, March 1978 and ANSI N510-19757 Or is this an error.

4.7.6.c.1 99.95% filter efficiency should most likely be,"in place penetration and bypass leakage type test."

4.8.1.1.1b SAFETY CONCERN-Reinstate this surveillance. The applicant has proposed deletion on the basis that this does not apply because the plant does not have a transfer feature for the 1E bus. This specification was written to test all of the bus transfers that interface between the offsite transmission circuit and the onsite 1E distribution system. This will i still leave seven bus transfers that '

i must be demonstrated as operable independent circuits.

4.8.1.1.2c & d SAFETY CONCERN- The applicant should add to the surveillance the requirement to I compensate for the fact that fuel oil will be stored in contact with inorganic zinc coating. See GPC letter dated June 4, 1986 in response to module 13B coatings regarding NRC IEC No. 77-15.

This is a site specific item.

4.8.1.1.2.5 Clarification needed- Is the loading of the D/G supposed to be from the time they start to load or from the time a start signal is received. It appears that this test should be integrated into TS 4.8.1.2.4.

10

i 4.8.1.1.3f

• 11.5 seconds to get to 484 RPM seems

'. slow when compared to 4.8.1.1.3g.4b page 3/4 8-6 which requires energization of the 1E bus within 10 seconds. Normal RPM i is 450.

3.8.2.1 This action statement sets no time frame for complete restoration.(i.e. If a battery bank is inoperable, then all I need is a charger to remain at power forever.) Suggest that the wording be:

a) With one of the hbove D.C. electrical sources inoperable, restore the inoperable D.C. electrical source to operable status within two hours or be in at least...

b) Immediately verify both chargers are operable and inservice within two hours and restore the battery bank to operable status within four hours, or be in at least.... This will allow time to restore the battery by taking advantage of the second charger availability.

4.8.2.1.e.4 Change "The battery charger" to "Each battery charger" since there is more than one per battery bank.

Table 4.8-2 The float voltages and specific gravities are not appropriate. The float

. voltages should be:

Cat. A Cat. B Allowable 12.20 12.20 12.17 specific gravity.

The nominal specific gravity of these batteries is 1.210, therefore these values appear to be too low (i.e. Cat B allowable being 11.190 is 20 points below nominal which indicates a discharged condition, not an operable condition.

l 3.8.2.2 & 4.8.2.2 Refer to action statement problems regarding TS 3.8.2.1.

3.8.3.1 Need to remove brackets in LCO or delete extra wording.

Il l . _ - . . _ . _ . _ .

J 4

3.8.4.2'. ,

Specify the FSAR table where these

. valves are listed, either in the, specification or the basis.

4.9.8.1 A surveillance requirement should be added to ensure that the required RHR loop shall be demonstrated operable pursuant to specif-ication 4.0.5.

4.9.8.2 A surveillance requirement should be added to ensure that the required RHR loop shall be demonstrated operable pursuant to specification 4.0.5.

4.9.12 Surveillance requirements need to reflect the snme requirements as the other filter systems. Too many blanks in this surveillance to evaluate. It is unacceptable as is.

4.10.4.3 S/G wide range level should be 18.5%.

Reference WCAP 11269- setpoint study.

3.11.1.1 & 3.11.2.1 SAFETY CONCERN- Since the TS has been modified to no longer require a plant shutdown if action a cannot be satisfied (i.e. immediately), then additional actions should be specified. This specification had self-regulation when tied to the reactor as the ultimate source of all radioactive material. It is recommended that action b be deleted unless other appropriate actions are j specified.

3.11.1.2 Action a. Are there drinking water supplies within 3 miles downstream of plant discharge? If so, then 40 CFR 141 would apply.

4.11.2.4 m Add redundant surveillance that has been

• deleted (STS 4.11.2.4.2) Note that TS 4.11.1.3.2 has remained. This is a similar issue.

l l

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t 4.11.2.5

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Reference should be to Table 3.3-10 of

'. Specification 3.3.3.10, not Table 3.3-13 of Specification 3.3.3.11 as shown on other page.

4.11.2.6.1 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> surveillance requirement has been changed to 7 days which appears to

- - - be an unreasonably long period of time for a sampl,e interval.

4.11.3b Typo- the process control program is in specification 6.12, not 6.13.

Table 3.12-1 Table notation 9 has a typographical error in that I-131 is referred to as V-131.

Table 3.12-2 The reporting level for activities are discrepant with the STS for a number of samples, all non-conservative. THese are: Zr-95 in water should be 300 instead of 400; Nb-95 in water should be 400 instead of 600; and La-140 in milk should be 300 instead of 400. Similarly, the lower limits of detection given are also non-conservative. These are: Zr-95 in water should be 15 instead of 30; Ba-140 in water should be 15 instead of 60; and Ba-140 in milk should be 15 instead of 60.

Bases 3/4 3.3.7 RG 1.95, dated Feb. 1975 should be RG 1.95, Rev. 1, Jan. 1977.This is in STS 1

and the FSAR section 1.9.9.5.

i 3/4 3.3.8 Have bases, but no specification. Add specification.

3/4 Cold Overpressure See comment for TS 3/4.4.9.3. Bases should clarify how the other two vent paths afford protection.

3/4 6.2.3 Containment Cooling Systems outage times have not been adjusted as stated. They are the same as STS which considered the system redundant, not overlapping.

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b FSAR section 1.9.7 states that

^

3/4 6.4 conformance is to RG 1.27 Rev. 2 Jan.

1986, not RG 1.27, Mar. 1974 as in the bases. -

Page B 3/4 8-1 Last paragraph, FSAR section 1.9.9 states conformance to RG 1.9, Rev. 2.

Dec. 1979, not to RG 1.9, Mar. 1971 as stated in bases.

Page B 3/4 8-2 Note comment for Table 4.8-2, page 3/4 8-13. Also IEEE Std. 450-1980 to read IEEE Std. 450-1975. FSAR section 1.9.129 states conformance. Add "IEEE Std." prior to "484-1975".

Page B 3/4 9-1 Remove [ ] from aroud "[2000]" as this is the value.

Page B 3/4 11.4 Typo- 5th line down, " Revision I" should read " Revision 1".

9 3/4 11.3 Delete "10CFR 50.36a" or clarify. This

= ~ - section of the code requires the submittal of the proposed technical specifications and thus should not be listed.

Design Features 5.3-1 The end of the second sentence is missing, "and contain a maximum total weight of 1766 grams uranium."

5.6.1.1 Conservative allowance of [2.6]% delta K/K for uncertainties needs to be completed.

5.6.3 Capacity of fuel storage. It should be resolved how this will be maintained since these are temporary racks. In order to install the permanent racks 14

these temporary racks will be removed

'. leaving no st 9rage capacity, .and replaced with an even greater capacity.

It is important to ensure that the applicant will have the ability to off-load a complete core.

- Table 5.7-1 Design Cycle- The Table does not reference the reactor vessel head vents.

The system is designed for only one cycle at power.

Administrative Controls .

6.1.2 This item deviates from the requirement to have a corporate officer, not the GMVNO perform this. See TMI item I.C.3 and I.A.l.2.

6.2.2e This specification has reduced GL 82-12 requirements in that:

1) Only overtime that exceeds the guidelines is reviewed. The STS has individual overtime reviewed monthly N which is the only means of monitoring 1

- gY the 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> objective.

p 2) The level of authorization is too low for the approval of excess overtime, it resides at the same level that approves regular overtime.

6.7.2 First line- add reference as to who reviews the procedures. This specification as written only requires periodic review (see last line). See below Suggest that 6.7.2 read:

Each procedure of 6.7.1 above, and changes thereto, Shall be reviewed and approved by the GMVNO or the department head of the responsible department, and the PRB, as delineated in 6.4.1.6 and 6.4.1.7a, prior to implementation and reviewed periodically as set forth in administrative procedures. The GMVNO shall review and approve the following procedures:...

6.7.3c 1) The " Reviewed in accordance with Specification 6.7.2" will only require periodic review.

2) The GMVNO does not review all l

changes. Suggest that 6.7.3c be changed 15 I

f .

,- t

~

to read ,"The change is documented, .

reviewed, and approved in accordance with section 6.7.2 within 14 days of implementation.

6.12.2b Clarify who accepts- suggest:"b shall

- become effective upon review and acceptance by the PRB and approval per Specification 6.7.2 4

6.13.2b Same as for 6.12.2b above.

General Comments

1) The TABLES AND FIGURES listed in the index should be indexed if fifteen spaces. For ease of readability.

. 2) Will TS 3.3.3.8, " LOOSE PART DETECTION SYSTEM" be in the TS or not.

If it is to be deleted, then TS 3.3.3.9 and higher need to be renumbered.

3) TS on page 3/4 6-20 should be listed as 3/4.6.5 vice 3/4.6.6.

4) The WESTINGHOUSE SETPOINT METHODOLOGY FOR PROTECTION SYSTEMS-VOGTLE STATION (WCAP 11269) needs to be incorporated into the Technical Specification Tables 2.2-1 and 3.3-3 to eliminate non-conservative values. Submitted to the NRC September 29, 1986.

Also affects TS's 4.4.1.2.2, 4.4.1.3.2 & 3.4.1.4.1b

5) Several Tables have been removed from the TS which is probably part

~

of a TS upgrade program. Proper references to where the information now resides must be made. If not directly as part of the TS, then in the bases section. Currently the bases sections have not been revised and read as if the tablos still exist.

6) In several action statements the words, " pursuant to Specification 6.9.2" appear. All actions need to be section 6.8.2 since this is the special report section.

16

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^'

b. The analyses take credit for reactor trip from the power range E ~. ' . ;"

(5:: - high neutron flux channels. The power range high neutron J.*r 1 '

U./ flux channels are not required to be operable during shutdown g

/y@ by the Technical Specifications. The source range and

~

intermediate range channels are required to be operable;'

however, the Technical Specification Basis (Page B2-4) s'tates that no credit was taken for these trips. Technical Specification Table 3.3-2 states that delay' times for the source and intermediate trip functions are not applicable.

Correct this apparent inconsistency between the Technical Specifications and the safety analysis as required by 10 CFR 50.36.

Response

In response to this concern and to the concerns raised in Part c. below, we are

- . preposing changes to Taoles 3.3-1 and 3.3-2 of the Technical Specifications wh ch will a) require two source range channels to be operable in modes 3-5 whenever the rods are energized and capable of being withdrawn and b) require response time testing for the source range channels. Under the revised tables, the source range trip will not be blocked during a startup until the power range channels are available. As such, the intermediate range trips are redundant and no credit.is taken for them in the safety analyses.

Table 3.3-2 of the Specifications will be revised to specify acceptable reactor trip system instrumentation response times of < 0.5 seconds for the source range 28 m

~ . -

__ c

{

y,a .,. e A Y.:. .'

fi'

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k:(', .s channels. This value is consistent with the safety analysis assumptions and with f/ - f 9jpeg . a gry . _. ; the existing specification for the power range channel. Note that for the case of

n w

. [* }"' a rod withdrawal from subcritical, several seconds would elapse between the time y, 4 I

when the sour _c_e range. trip s.igna.l is. y,enerated (when the core is ,%t or.just below critical) and the time when the protection is actually needed (significant power generation occurring)- see, for example Figures.15.2-1 to 15.2-3 in the 'JISAR.

The 0.5 second response time, which is used for consistency with with t'.se other channels, is therefore more than adequate to protect against rod withdrawal e/ents in modes 3-5. We have determined that response time testing of these channels can be implemented with minimal dif ficulty.

o G

O e e O

h 29

,g . vive 6 12::1 tc.c 3 c2-IpgMDRA) CCMMENIS ON V0GTLE LINIT 1 9900V)\MVWMD KC@KFLMECWUAE W v ' V EEP 5tCTION 7) t , / '-

If In Section 7.2.2.2 of the SER, a technical specification requirement for the turbine

• rip on reactor trip circuitry is discussed. Provide appropri-ste technict' specification. ,.

t 2 Proposed technical specifications include changes approved by the staff's myiew of WCAP-10271. The staff's letter of July 24,1985, and the " West- '

inghouse Owners Group Guidelines for Preparation of Submittals Requesting Revisions to RPS Technical Specifications" related to WCAP-10271 discuss conditions upon which approval of technical specification changes is granted.

Provide comitments covering programs or procedures which address comon cause problems and instrument setpoint drift.

o On page 2-7 under NOTE 1: Cormet the conflict between the equation used- G' here and that used on page 7.2.1-5 of the FSAR.

1. On page 2-g under NOTE 3: Correct the conflict between the equation used gf here and that used on page 7.2.1-6 of the FSAR.

o Sections 4.3.1.2 and 4.3.2.2 refer to limits on response times. Tables C

covering response times have been eliminated from the technical specifica-tions. Provide a reference to the source of response time limits in these -

two sections. # # # 8 dd#"" ^"" Y G

8 ..

4

a/isess 12$11 Id.cis cas 6/ On page 3/4 3-2 under Functional' Unit 5.b: Entry under " MINIMUM CHANNELS '

,,op m OPERABLE" has been reduced to one. Provide3 justification.

4 o on pages 3/4 3-g and 3/4 3-10: Functional Unit 18.d is missing. Renumber entries as appropriate. # "b' '~

On page 3/4 3-17 under Functional Unit 3.b.1: Rewrite entry around manual or initiation of containment spray and isolation. Under Functional Unit 4.a.1:

Correct the conflict between the logic' used here and that shown in Figure

+

7.2.1-1 (Sheet 8) of the FSAR. Under Functional Unit 3.b.5: Entries under

" MIN! MUM CHANNEL 5 OPERABLE" should be revised te encompass the redundancy ,

provided by the design, i

I o on page 3/4 3-18 under Functional Unit 5: Provide appropriate entries for y l

feedwater isolation on low Tavg coincident with P-4 asshown in Figure 7.2.1-1(Sheet 13)oftheFSAR.

8I

/o On page 3/4 3-20 under Functional Unit 10: Provide appropriate entries for control room isolation on high chlorine input as shown in Figure 7.2.1-1(Sheet 8)oftheFSAb.'Entryunde " ACTION" deviates from WesE d nghouse 5TS. Provide justification. , e

, Q p p ,r f , pyg yyy ,

,t s ssw"q wws-g .. ..

O

12/19 56 12. 2 NC.213 C04

-3

/ -

,o On page 3/4 3-16 under Functional Unit 3.a.4: Entry under " ACTION" devi-ates from Westinghouse STS. Provide justification. ( *** #5 /we f On pages 3/4 3-17 and 3/4 3-18 under Functional Unit 4: Reference to foot-note f deviates from Westinghouse STS. Provide justification.

/ is pH n s !/*** I pi In Tab ks 3.3-2, 3.3-3, and 4.3-2: Add instrumentation for all other ESFAS

,/

functions, such as control building ESF electr'ical equipment rooms HVAC, th

/ at are not included in other sections, ofpsto the technical specifications.

i pre. t=Afr, At8 r~ndJ

/

d on page 3/4 3-19 under Functional Unit 6: Provide appropriate entries for auxiliary feedwater pump suction transfer following the guidance of the Westinghouse STS. [W# '

i

/

i On page 3/4 3-26 under Functional Unit 4.e: Entries under " TRIP SETP0 INT" and " ALLOWABLE VALUE" should be positive following the guidance of the Westinghouse STS.

o On page 3/4 3-30 under " TABLE NOTATIONS:" Verify correctness of " s. g,". ..

for 72 in note *.

0 9

g .. ..

4

1

. 12/15/56 13:12 tc.213 225 4-

/ -

8 In Table 4.3-2: Entries under "5 LAVE RELAY TEST" should be changed to G#

~

4 "M" in lieu of "Q" per Section 7.3.3.3 of the staff's SER. Also, the additional relief prnvided by Enclosure 3 to the staff't. July 24,1985 letter related to WCAP-10271 shculd be considered. -

1

,b o In Table 3.3-4: Cross-references to other tables should be corrected.

1_ (su semd) l ,

On page 3/4 3-4g: Delete "[Illustrational only]" and verify that all 9 appropriate entries have been included in these tables.

• o ' In Table 3.3-7: Add appropriate entries and correct the table following . -

the Westinghouse STS and the plant design.

o On page 3/4 3 51 under Instrument 14: Expand entries to include wide range 5' and narrow range instruments following the plants' R.G.1.g7 Type A designa-tion for these instruments as discussed in this staff's SER.

e

.o In Table 3.3 8: Add appropriate entries for containment isolation valve d -

positions following the plant's R.G. 1.97 categorization of this instru-

/

mentation as Category 1.

g .. ..

(

. 12n956 13: 12 m. ais ces

-5. .

r

]

On page 3/4 3-68 under " ACTION": Clarify "less than required" if this ###

is intended to refer to "Minimsm Channels,.0perable."

7do On page 3/4 5-2 under Section 4.5.1.2: Required surveillance deviates c e from the Westinghouse STS. Provide justification., (/ Av ne e# 6* *"M

/o On page 3/4 6-21 under Section 4.6.6.d.2: Since the electrical penetration room exhaust air cleanup system is initiated automatically by more signals than just safety injection and surveillance requirements for the instrumenta-tion for those other signals are not now included under Table 4.3-2, include appropriate surveillance requirements or, as stated above, include the instrumentation under Tables 3.3-2, 3.3-3, and 4.3 2. /t** d5 '1>

/ Un page 3/4 7-11 under Sectlan 4.7.3.bt Since the component ecoling water dr o

< system pugs start is initiated automatically by more signals than just safety injection and survei11snce requirements for the instrumentation for '

those other signals are not now included under Table 4.3 2, include appro-priate surveillance requirements or, as stated above, include the instru-i mentation under Tables 3.3-2, 3.3-3, and 4.3-2.

. I On page 3/4 7-12 under section 4.7.4.b

Since the nuclear service cooling se water system pumps start is initiated automatically by more signals than g .. ..

t

,, 10/19/86 13:13 NO.013 007

.t k

just safety injection and surveillance requirements for the instrumentation for those other signals are not now included under Table 4.3-2, include appropriate surveillance requirements or, as stated above, include the f

instrumentation under Tables 3.3-2, 3.3-3, and 4.3-2.

th 7 Onpages3/47 hand 7-16: Entries under Section 4.7.6.e.2 and Section F, 4.7.6.e.5 apposr to be duplicate. Verify that separate entries are re-quired. I##I 7c, o On page 3/4 7-2g under Section 4.7.11.b and c: Since the ESF room cooler G ,

system is initiated by more signals than just safety injection and sur-veillance requirements for the instrumentation for those other signals am not now included under Table 4.3-2, include appropriate surveillance re-quirements or, as stated above include the instrumentation under Tables 3.3-2, 3.3-3, and 4.3 2.

a In Table 3.311 under " Electric Steam Boiler Isolation": Correct the con

• r '-

f11ct between the entries under " Instrument Channel" and Figure 7.6.6-6 of the FSAR.

! o On page B 3/4 3-1: Include in the second paragraph a reference to the S#

v staff's February 21,1985, SER on WCAP-10271. Also in the fourth line j from the bottom insert a "+" in the equation. ,, ,,

0

10/19/G6 13:14 NO.013 008

/

On page 8 3/4 3-3: Correct the discussion provided under "P-11" to agree f<*

with that provided in Table 7.3.1-3 of the<FSAR..

' 4

..e In Section 6.7: Add programatic requirements for surveillance and controls for restoration of inoperable instruments for all plant post-accident non-itoring instrumentation that are classified Category 2 or 3 per R.G.1.97 following the guidance provided in an October 12,1983, memo from Roger J.

Mattson to Darrell G. Eisenhut covering technical specifications for post-accident monitoring instrumentation.

e.. g/ In Tables 2.2-1 and 3.3-3: Correct entries under " TRIP SETP0 INT" and

" ALLOWABLE VALUE" to agree with those pmvided in WCAP-11269, " Westinghouse 5etpoint Methodology For Protection Systems Vogtle Station."

In WCAP-11269 errors are included for Veritrak pressure transmitters used ('

/G In light of Westinghouse's recent notifi-

! in several protection channels.

cation to licensees concerning possible excessive errors in these trans-mitters, affected channels should be identified and notes included in the l

technical specifications to ensure appropriate interim measures are taken.

gge on page 3/4 3-12 under " TABLE NOTATION $" hote 11 should be made consistent f.0 with corresponding note 12 of Generic Letter 86-09 to require ' independent testing of the undervoltage and shunt trip attachments of the' reactor trip a .. ..

breakers for each train every 62 days on a staggered test basis.

4 4

4

.. lacs 2:14 als , es g g

\/o GTL 5 T S 's QNW ma umrw ps.- EM

1. I:ngineered safety Features Actuation System Instroentation. Table 3.3-2 LPage3/43-15)

,Wtomat_ic Safety In,1ectin_n (hish contaiment'pressurgl is not reevired to 1

Cao-raMe in w+t 4 ty TabTeT. 42. In a letidFfrom J. Bailey grc to 4

1. Denton NRC Decchiber 9 1385 an analysis of large break LOCA in Mode 4 mas provided to the staff assum;ing ispediate and automatic actuation of 5! et the end of blowdown. Provide revisions to either.the Safety Analysis or the Technical Specifications so that they are consistent. If.

you choose to avise the Safety Analysis the operator response time to i

annually actuate 51 should be Justifled. In other reviews the staff has 5 secepted operator response times of 10 minutes following a control room a1erm. . . ~ .

2. Seactor Trip System and En Instrumentation (page S432) 3/gineered Safety Features Actuation fEystem ")

i The bases descNbe/tNe japortance of response time testing for the v

, ReacDr Trip and pgineered Safety Features actuation functions. I Limiting conditions for operation are not provided in the Technical

! Specifications for the resoonse 11asa er for their surveillance. ple e .

4errect this teconsistency.

3. . Rosetor The FSAR Trip Instrumentation.

Evaluction Tablecontrol of inadvertant 3.31 (page 3/4 3 2) l from bank withdrawa subcritical or low power assumes reactor trip to be initiated b

' neutron flux signal from the power range channels (Iow setting)y a high

. The - '

i Technin1 #Meit_ications do not require th tporar range thanne n to be '

reager 3 Mitful~laodes 3.e . ena ps. le esse; og_eJ.aD. .e. w wn correct this econsistency. f pur reiponse is that the reactor range would I

! trip frca signals generated by tw source omdatwtumettets channels. Provide the response times for this instrve ntation under ites 3 and demonstrate that the transient analysis in the F5AR is bounding. s,

4. Reactor Coolant System Not Shutdown 3.4.1.3 and $1d %tdown 3.4.h and i 3.4.1.4.2(pages3/44-3to3/44-5)

The F5AR evaluction for inadvertent control bank withdrawal free suberitical asswees that two reactor coolant pumps are operating. The Technical Srcification do not require any reacter coolant _ pumps _to_be in gpera", fen Tn mo&r a m e w tr a smaroi roo manrwit6drawal transient were no occur without coolant pump flows The minimum DNBR afght be decreased below that calculated in the 75AR. provide additional safety analyses of inadvertent control red withdrawal transients in undes 4 and l 5 without reactor coolant pump flow or demonstrate that inadvertent criticality from control red bank withdrawal cannot occur in Modes 4 and 5. ,

I i l l

~

10/15/56 13:15 NO.013 010 2

5. 8t _

' sin steam Line Isolation Valves 7 J (Page 3/4 7 9) and Engineeree Safety 3-17) Features Actuation Systems natraentations Table).3,2 fpage 3/4 The Safety Evaluation in the FSAR for steam generator tube ptureru...

MSIV on the associsted steam line. The Technicel Specifiert require manuel isolatoen ennahtiti er awreen ons do not ,

tv af the er va a unde

- rrovles assistonal safety ana yses o" an unisolatable steam generator eenststenture accident in Mode 4 er provide Technical Speciffeations the avrrent Safety Analysts. *

'8. Steam Line Atmospheric Relief valves 2, The FSAR Safety Evaluation of Steam Generator Tube Ruptute assumes operator action te open the atmospheric relief valves en the steam lines

• of the unaffected steam generators. This action is required te limit rediation release to the atmosphere if offsite power is test. The Inghnical_Specificatu evi ur u m t. Provide ans da est require operab111_tv of the,atnesoberic ad51ticrci at.Ts.v snavne sW a steam generator tube rupture accident with inoperable steam line relief valves er provide Technical Speciffections . .. - that are consistent wf th the safety Analysis.

7. ,

Special Test taesptish 3/4.10.4 Reactor Coolant Loops (page '

power Operation in Mode 1 is permitted to the p.7 interlock. 3/4.10.4) setpoint which may be in excess of 105 power.

The 75AR does not evaluate power

,- operation vithout the vesctor ecolant pumps. Either provide a supporting Safety Evaluatten er revise

'g {contstentwiththeexisting' the Technical safety'ahalysis. ~

Specifications so that they are

~

l l

l l

1

huM+;M47 ~

INSTRUMENTATION Y 3/4.3.4 TURBINE OVERSPEED PROTECTION ,

• LIMITING CONDITION FOR OPERATION 3.3.4 At least one Turbine Overspeed Protection System shall be OPERABLE..

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With one stop valve or one control valve per high pressure turbine steam line inoperable and/or with one intermediate stop valve or one intercept valve per low pressure turbine steam line inoperable, restore the inoperable valve (s) to OPERA 3LE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or close at least one valve in the affected steam line(s) or isolate the turbine from the steam supply within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. .

b. With the above required Turbine Overspeed Protection System otherwise inoperable, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> isolate the turbine from the steam supply.

SURVEILLANCE REQUIREMENTS 4.3.4.1 The provisions of Specification 4.0.4 are not applicable.

4.3.4.2 The above required Turbine Overspeed Protection System shall be (h

demonstrated OPERABLE: ,.

a. At least once per 7 days while in MODE I and while in MODE 2 with the turbine operating, by cycling each of the following valves through at least one complete cycle from the running position:

1) Four high pressure turbine stop valves, 2 57 MJ, r. :::r: t cM x s nt. m '. 4,,

R/) Six low pressure turbine intermediate stop valves, and 3/) Six low pressure turbine intercept valves.

b. At least once per 31 days while in MODE I and while in MODE 2 with the turbine operating, by direct observation of the movement of each of the above valves4through one complete cycle from the running E5 "' and Oc [ur byh getSSur b be to d */

• b S,

c. At least once per 18 months by performance of a CHANNEL CALIBRATION on the Turbine Overspeed Protection Systems, and

' d. At least once per 40 months by disassembling at least one of each of the above valves and performing a visual and surface inspection of valve seats, disks and stems and verifying no unacceptable flaws or

., Corrosion. 1

()At.!udlO h(. 8H4 f i

{ ff 5I RfI-r ng g40 (D CATAWBA - UNIT 1 3/4 3-91 Vklv'st)

.