ML20212D817

Summary of 861020-24 Meetings W/Utils,Bechtel & Westinghouse Re Tech Specs.Util Meeting Notes & Handouts Encl

ML20212D817
Person / Time
Site:	Vogtle
Issue date:	12/22/1986
From:	Mark Miller Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8701050047
Download: ML20212D817 (23)

Text

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Docket No.: 50-424 APPLICANT:

Georgia Power Company gEQSSN FACILITY:

Vogtle, Unit 1

SUBJECT:

Sufif4ARY OF TECHNICAL SPECIFICATI0flf1EETING HELD OCTOBER 20-24, 1986 The staff ret with the aoplicant and its representatives between October 20 through 24, 1986, to discuss the Vogtle Unit 1 Technical Specifications.

Participants are listed in Enclosure 1.

By letter cated December 5,1986, the applicant foniarded its meeting notes and handouts from this aceting. The meeting notes are included as Enclosure

2. contains the meeting handouts.

Melanie A. f1 iller, Project Manager PWR Project Directorate #4 Division of PWR Licensing-A

Enclosures:

As stated cc: See next page ll If

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MEETING

SUMMARY

DISTRIBUTION b

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NRC Participants NRC PDR

5. Brown L PDR J. Thompson

.NSIC-C. Moon

.PRC Systen C. Willis PWR#4 Reading File P. Moore Project Manager M. Miller F. Burrows M. Duncan J. Rogge OGC W. Jensen J. Partlow D. Ilickman E. Jordan L. Crocker B. Griraes S. Chan ACRS (10)

T. Sullivan J. Pulsipher D. Tondi W. Swenson J. Lazevnick OTHERS R. Giardina C. Li J. Wing bec:

Licensee & Service List

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Mr. J. P. 0Reilly Georgia Power Company Vogtle Electric Generating Plant cc:

Mr. L. T. Gucwa Resident Inspector Chief Nuclear Engineer Nuclear Regulatory Commission Georgia Power Company P. O. Box 572 P.O. Box 4545 Waynesboro, Georgia 30830 Atlanta, Georgia 30302 Mr. Ruble A. Thomas Depoish Kirkland, III, Counsel Vice President - Licensing Office of the Consumers' Utility Vogtle Project Council Georgia Power Company /

Suite 225 Southern Company Services. Inc.

32 Peachtree Street, N.U.

P.O. Box 2625 Atlanta, Georgia 30303 Birmingham, Alabama 35202 James E. Joiner Mr. Donald O. Foster Troutman, Sanders, Lockerman, Vice President & Project General Manager

& Ashmore Georgia Power Company Candler Buildine Post Office Box 299A, Route 2 127 Peachtree Street, N.E.

Waynesboro, Georgia 30830 Atlanta, Georgia 30303 Danny Feig Mr. J. A. Bailey 1130 Alta Avenue Project Licensing Manager Atlanta, Georgia 30307 Southern Company Services, Inc.

P.O. Box 2625 Carol Stangler Birminghem, Alabama 35202 Georgians Against huclear Energy 425 Euclid Terrace Errest L. Clake, Jr.

Atlanta, Georgia 30307 Bruce W. Churchill, Esq.

Shaw, Pittman, Potts and Trowbridge 2300 N Street, N.W.

k'ashington, D. C.

20037 Fr..G. Bockhold, Jr.

Vogtle Plant Manager Georgia Power Company Route 2, Box 299-A

.Waynesboro, Georgia 30830 Regional Adninistrator, Region II U.S. Nuclear Regulatory Commission 101 Marietta Street, N.W., Suite 2900 Atlanta, Georgia 30323 Mr. R. E. Corway Senior Vice President and Project Director i

Georgia Power Company Rt. 2, P. O. Box 299A Waynesboro, Georgia 30830

4 4

ENCLOSURE 1 Participants NRC BECHTEL S. Brown S. Mahler J. Thompson S. Cereghino C.lbon C. Willis GEORGIA POWER COMPANY P. Moore F. Burrows J. Hartka J. Rogge R. Hand W. Jensen W. Kitchens D. Hickman D. Hudson L. Crocker P. Kochery S. Chan E. Cobb T. Sullivan J. Pulsipher WESTINGHOUSE D. Tondi W. Swenson E. Burns J. Lazevnick R. Morrison R. Giardine K. Daschke C. Li J. Wing SOUTHERN COMPANY SERVICES J. Stringfellow I

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4 Meeting Notes These notes reflect a page-by-page review of the Proof and Review copy.

The order in which comments were discussed was dictated largely by the availability of the branch reviewers throughout the week. These notes are organized in the order in which each comment was discussed.

1.

Pages 3/4 3-17 and 3-21: Our proposed revisions to Functional Unit 3b and our proposed footnote g were acceptable to the staff.

2.

Page 3/4 3-25:

Proposed revisions were acceptable.

3.

Pages 3/4 3-32 and 3-33: Proposed revisions were acceptable.

4.

Tables 3.3-2, 3.3-3 and 4.3-2: We agreed to add the ventilation radiation monitors which actuate the fuel handling building post accident ventilation system to these tables.

5.

Page 3/4 3-38:

Proposed revisions were acceptable.

6.

Page 3/4 3-53: The staff will have to give our proposed revisions further consideration.

7.

Table 3.3-3: We will have to provide radiation monitor setpoints for this table.

(These setpoints were provided on November 14, 1986, GN-1184 ).

8.

Page 3/4 9-14: We agreed to provide a specification on the fuel handling building post-accident ventilation system.

9.

Tables 4.3-5 and 4.3-6:

We proposed to delete the requirement for an analog channel operational test for flow rate measurement devices or monitors on the basis that these instruments do not provide alarm, trip or interlock functions. The staff found this acceptable.

10. Table 3.3-9, Action Statement 36b: The staff will have to give our proposed revision further consideration.

11. Table 4.3-5, Table Notations 1 and 2:

Proposed revisions were acceptable.

12.

Page 3/4 3-60:

Proposed revision to 3.ll.2.1.a was acceptable.

13. Pages 3/4 7-15 and 7-18: The staff took the position that surveillance requirements c.2 and d should specify 30*C as opposed to 80*C.

We agreed to revise the requirements to 30*C but GPC will check to see if this causes any difficulties.

I

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Attachment.1.,

Pase 2 14.

Pages 3/4 3-62, 3-65, 3-66, 3-67:

Proposed revisions to operability requirements on samplers / monitors were acceptable.

15.

Pages 3/4 6-20 and 6-21: The staff agreed to delete this specification on the basis that the FSAR will be revised so that the electrical penetration room filtration system is no longer referred to as an ESF system.

16.

Page 2/4 4-28:

The NRC staff stated that the definition of a gross radioactivity determination as it appears on this page states exactly what they expect to be done. The refore, the revisions proposed by GPC were not acceptable. However, the staff did indicate that they would entertain some clarification in the bases. We proposed clarifying statements for pages B 3/4 4-7 and B 3/4 7-2 for consideration by the staff.

(The proposed revisions are reflected in our November 6,1986 submittal, GN-1166.)

17.

Table 4.7-1:

The staff will consider the deletion of the requirement to do a gross radioactivity determination and require only the DEI determination once per 31 days.

18.

Page 3/4 7-14:

Our proposed deletion of the surveillance requirement on control room air temperature at 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> intervals was not acceptable.

We had proposed the deletion on the basis that the emergency system does not maintain control room air temperature during normal operation. In turn, we would perform surveillance on a monthly basis that would demonstrate the capability of the emergency system to maintain control room air tempe rature.

However one of the members of the regional staff expressed concern that if control room temperature rose too high during normal operation, the emergency system might not be able to maintain temperature l

if called upon. GPC chose to drop the proposed revision rather than pursue an answer to this question.

19.

Pages 3/4 7-15, 7-17 and 7-18: The staff would not accept our argument for an ef ficiency of 99.5% as opposed to 99.95% on the basis that 99.95%

was " traditional."

20.

Page 3/4' 7-16: We must provide the correct pressurization flow rate.

(This was provided in our November 14, 1986 submittal, GN-1184.)

21.

Page 3/4 7-18:

Item d.3 was revised to require a negative pressure of 1/4 inch relative to the outside atmosphere.

22.

Page 3/4 7-14:

The NRC staff will have to reconsider the applicability which appears in the Proof and Review Copy for consistency with Specification 3.3.3.7.

O o

Attachment-i- -

Pese 3 23.

Pages 3/4 11-1 and 11-5: The staff did not accept our proposal to put Tables 4.11-1 and 4.11-2 into the FSAR. Their position was that if the tables are to be removed from the Technical Specifications, they should be relocated to the ODCM. GPC did not wish to include the tables in the ODCM and therefore the tables will remain in the Technical Specifications.

24.

Pages 3/4 11-1 and 11-5: Both of these specifications include exceptions to the provisions of Specifications 3.0.3 and 3.0.4.

The regional staff commented that, from the standpoint of enforceability, these exceptions should not apply since forcing a shutdown will definitely call to utility management's attention any noncompliance with these specifications. We responded that forcing plant shutdown, especially in the case of gaseous effluents, will only serve to exacerbate the situation.

If the plant was found to be operating ou' side the limits of these specifications and appropriate action had not been taken to return to compliance with the limits, then the utility is subject to enforcement action regardless of whether or not the plant is shut down.

25.

Pages 3/4 12-5, 12-6, 12-9, and 12-13:

Proposed revisions were found to be acceptable.

26.

Pages B 3/4 11-4 and 11-5:

Proposed revisions were found to be acceptable.

27.

Page B 3/4 12-1: Our proposed revisions to these pages were not accepted.

28. Tables 3.12-2 and 4.12-1: The staff questioned the limits we proposed for Zr-95, Nb-95, Ba-140, and La-140. We provided the staff with further information on the basis for our proposed limits.

29.

Page 6-8, item 1: The staff questioned our proposed limits for PRB review of accidental, unplanned releases. Our proposed revision was not incorporated into the final draf t.

We did however provide the basis for our proposal in our November 14, 1986 transmittal CN-1184.

30.

Pages 6-18 and 6-19:

Proposed revisions were found to be acceptable.

31.

Page 6-14: Our proposed revision was found to be acceptable but there is a mistake on page 11-2 of Supplement 3 to the SER which must he corrected. The SER reads " Containment Spray system including the NaOH subsystem" and it'should read " excluding the NaOH subsystem."

32. Table 4.3-1:

The staff's position is that reactor trip breaker bypass breaker testing will remain a tech. spec. requirement.

Attachment.1-Page 4 33.

Page 3/4 3-50: We agreed to list all channels which are classified as j

Category 1 as defined in Regulatory Guide 1.97.

Our proposed Action Statements were then revised to address those cases where we had an =

channels of instrumentation than required so the extra channels ara n >t required to be operable by the tech. specs. The staff also took the position that containment isolation valve position indication be classified as Category 1.

Our FSAR classifies containment isolation position indication as Category 2 on the basis that only single position indication is provided. This is the only deviation from Category 1 criteria. The staff found this acceptable in the SER and maintained that this instrumentation should be treated as Category 1 in the tech. specs.

As a result, the final draft will reflect containment isolation valve position indication as Category 1 instrumentation.

At this point in the meeting we began a discussion of the ICSB reviewer's In the subsequent notes the ICSB reviewers comments will appear comments.

followed by the disposition of the comment.

34.

In Section 7.2.2.2 of the SER, a technical specification requirement for the turbine trip on reactor trip circuitry is discussed. Provide appropriate technical specification.

Response: Turbine trip on reactor trip is accomplished by interlock P-4 which is listed in Tables 3.3-2, 3.3-3 and 4.3-2 as Functional unit 9b.

35.

Proposed technical specifications include changes approved by the staff's review of WCAP-10271. The staff's letter of July 24, 1985, and the

" Westinghouse 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 commitments covering programs or procedures which address common cause problems and instrument setpoint drift.

Response: GPC agreed to provide this commitment in the meeting notes which will in turn be provided to the staff as documentation of this meeting.

Plant Vogtle administrative procedures and engineering procedures are under revision to include the requirement of common mode failure analysis in the event that instrumentation fails the Reactor Protection System 3 months Staggered Base Surveillance Test.

If a common mode failure is identified, the corrective action is included in the requirements for retesting the redundant protection channels.

The Instrumentation and Controls Department has developed a program to trend setpoint drif t.

This program will apply to instruments currently under consideration by the NRC staff regarding a changa in analog channel operational test frequency provided by WCAP 10271. A computer program

o Pase 5 has been established to facilitate the trending, analysis and graphing of the associated analog channel operational test results. The loops that must be trenced have been identified and a list has been forwarded to the appropriate personnel. Copies of the completed and approved analog channel operational test data sheets will be provided for periodic computer data base update. The trending results/ graphs can be displayed or printed out upon request. Hard copies of the trending results/ graphs will be maintained in a central file along with the copies of the approved analog channel operational test data. A procedure is currently being draf ted to fully document this Setpoint Drif t Trending Program.

36.

On page 2-7 under NOTE 1:

Correct the conflict between the equation used here and that used on page 7.2.1-5 of the FSAR.

On page 2-9 under NOTE 3:

Correct the conflict between the equation used here and that used on page 7.2.1-6 of the FSAR.

Response: The FSAR is not actually in conflict with the tech specs.

However, we agreed to revise the FSAR to provide clarification.

(A mark-up of the FSAR pages in question was provided in our November 17, 1986 submittal, GN-1189).

37.

Sections 4.3.1.2 and 4.3.2.2 refer to limits on response times. Tables covering response times have been eliminated from the technical specifications.

Provide a reference to the source of response time limits in these two sections.

Response: We provided the requested reference in the bases for 3/4.3.1.2 and 3/4.3.2.2 in our November 6,1986 submittal, GN-1166.

38. On page 3/4 3-2 under Functional Unit 5.b:

Entry under " MINIMUM CHANNELS OPERABLE" has been reduced to one.

Provide additional justification.

Response: The ICSB reviewer stated thac if we could resolve this with the RSB reviewer then he would be satisfied. We discussed our justification provided in our February 28, 1986 submittal with the RSB reviewer and our proposed revision was found to be acceptable.

39.

On pages 3/4 3-9 and 3/4 3-10:

Functional Unit 18.d is missing.

Renumber entries as appropriate.

Response

This discrepancy was corrected in our October 13, 1986 submittal, GN-1115.

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

Under Functional Unit 4.a.It 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 " MINIMUM CHANNELS OPERABLE" should be revised to encompass the redundancy provided by the design.

6 Attachment.1_ _,

Page 6 Response: The entry for manual initiation of containment spray and containment ventilation isolation was left as is in the Proof and Review Copy. With regard to Figure 7.2.1-1 (Sheet 8) and manual steamline isolation, we agreed to correct the FSAR figure to agree with the tech.

specs. The comment about Functional Unit 3.b.5 was resolved with our proposed revision.

41.

In WCAP-11269 errors are included for Veritrak pressure transmitters used in several protection channels. In light of Westinghouse's recent notification to licensees concerning possible excessive errors in these transmitters, affected channels should be identified and notes included in the technical specifications to ensure appropriate interim measures are taken.

Responses: We agreed to the more conservative setpoints recommended by Westinghouse as interim corrective action until the problem is finally resolved.

At this point, we deferred discussion of the remainder of the ICSB reviewer's comments and moved to the RSB reviewer's concerns.

42.

Engineering Safety Features Actuation System Instrumentation, Table 3.3-2 (Page 3/4 3-15): Automatic Safety Injection (high containment pressure) is not required to be operable in Mode 4 by Table 3.3-2.

In a letter from J. Bailey, GPC, to H. Denton, NRC, December 9, 1985, an analysis of large break LOCA in Mode 4 was provided to the staff assuming immediate and automatic actuation of SI at the end of blowdown.

Provide revisions to either the Safety Analysis or the Technical Specifications so that they are consistent. If you choose to revise the Safety Analysis the operator response time to manually actuate SI should be justified.

In other reviews the staff has accepted operator response times of 10 minutes following a control room alarm.

Response: We resolved this concern by agreeing to put operability requirements on safety injection initiation due to high containment pressure in Mode 4.

43.

Reactor Trip System and Engineered Safety Features Actuation System Instrumentation (Page B 3/4 3-2)

The bases describe the importance of response times testing for the Reactor Trip and Engineered Safety Features actuation functions.

Limiting conditions for operation are not provided in the Technical Specifications for the response times or for their surveillance. Please correct this inconsistency.

l ^ ~'

Pase 7 Response: We resolved this item by providing a reference for response time limits in the bases as discussed with the ICSB reviewer.

44.

Reactor Trip Instrumentation, Table 3.3-1 (Page 3/4 3-2): The FSAR Evaluation of inadvertent control bank withdrawal from suberitical or low power assumes reactor trip to be initiated by a high neutron flur signal from the power range channels (low setting). The Technical Specifications do not require the power range channels to be operable when the reactor is suberitical (modes 3, 4, and 5).

Please correct this inconsistency.

If your response is that the reactor would trip from signals generated by the source range channels. Provide the response times for this instrumentation under item 2 and demonstrate that the transient analysis in the FSAR is bounding.

Reactor Coolant System Hot Shutdown 3.4.1.3 and Cold shutdown 3.4.4.1 and 3.4.1.4.2 (Pages 3/4 4-3 to 3/4 4-6):

The FSAR evaluation for inadvertent control bank withdrawal from suberitical assumes that two reactor coolant pumps are operating. The Technical Specification do not require any reactor coolant pumps to be in operation in modes 4 and 5.

If a control rod bank withdrawal transient were to occur without coolant pump flow, the minimum DNBR might be decreased below that calculated in the FSAR.

Provide additional safety analyses of inadvertent control rod withdrawal transients in modec 4 and 5 without reactor coolant pump flow or demonstrate that inadvertent criticality from control rod bank withdrawal cannot occur in modes 4 and 5.

Response: Both of these concerns were resolved by a commitment to perform response times testing on the source range instrumentation and adding a statement to the bases for Section 2.0 to the effect that the source range instrumentation will be used to mitigate the consequences of an inadvertent control rod bank withdrawal in Modes 3, 4 and 5.

45.

Main Steam Line Isolation Valves 3.7.1.5 (Page 3/4 7-9) and Engineered Safety Features Actuation Systems Instrumentations Table 3.3-2 (Page 3/4 3-17)

The Safety Evaluation in the FSAR for steam generator tube rupture assumes that the operator takes action to isolate the leak by closing the MSIV on the associated steam line.

The Technical Specifications do not require manual isolation capability or operability of the MSIVs in Mode 3.

Provide additional safety analyses of an unisolatable steam generator tube rupture accident in Mode 4 or provide Technical Specifications consistent with the current Safety Analysis.

l Attachment _i_ _,

Pese 8

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Steam Line Atmospheric Relief Valves The FSAR Safety Evaluation of Steam Generator Tube Rupture assumes operator action to open the atmospheric relief valves on the steam lines of the unaffected steam generators. This action is required to limit radiation release to the atmosphere if offsite power is lost.

The 4

Technical Specifications do not require operability of the atmospheric relief valves.

Provide additional safety analyses of a steam generator tube rupture accident with inoperable stam line relief valves or provide Technical Specifications that are consistent with the Safety Analysis Response: Both of these items were resolved by the fact that we are consitted to a license condition which requires us to follow the 4

recommendations resulting from interaction between the NRC staff and WOG Steam Generator Tube Rupture Subgroup.

46.

Special Test Exception 3/4.10.4 Reactor Coolant Loops (Page 3/4.10.4)

Power Operation in Mode 1 is permitted to the P-7 interlock setpoint which may be in excess of 10% power. The FSAR does not evaluate power operation without the reactor coolant pumps. Either provide a supporting Safety Evaluation or revise the Technical Specifications so that they are consistent with the existing safety analysis..

Response: We vill have to investigate this item and provide the staff with additional information.

47.

Pages 3/4 1-7, 1-8, 1-9, 1-17, and 2-8: Our proposed revisions were found to be acceptable by the NRC staff.

48 Page 3/4 2-14:

The staff accepted our proposed revision concerning the 12-hour surveillance requirement on RCS flow rate.

49.

Pages 3/4 3-2 and 3-11:

Proposed revisions were acceptable.

50.

Page 3/4 3-41: The staff will have to give our proposed revisions further consideration.

51.

Page 3/4 3-50:

This spec was revised as noted in our November 6,1986 submittal (GN-1166). The staff will have to give our proposed Action 32 further consideration.

52.

Page 3/4 4-7:

The staff found our proposed revision to be unacceptable on the basis that the COMS was not reviewed as a safety grade system.

i, However the staff offered an alternative which allows the pressurizer code safeties to be inoperable in Mode 5 if an equivalent size hole is provided. GPC found this to be acceptable.

1 53.

Page 3/4 4-10:

This spec was revised as noted in our November 6,1985 submittal (GN-1166).

54.

Page 3/4 4-6: Our proposed revisions were found to be acceptable.

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Pase 9 55.

Page 3/4 5-4:

Proposed revisions were acceptable provided the FSAR is revised to reflect that valves HV 8803 A&B are now effectively locked in position.

56.

Pages 6-1 and 6-5, footnote

• and Table 6.2-1: Our proposed revision was not accepted by the staff.

57.

Figures 6.2-1 and 6.2-2: We had proposed to delete these figures on the basis that the figurec were difficult to keep up-to-date on the basis that each change in organization would require a tech spec amendment.

The offsite and onsite organizations are documentcd in Chapter 13 of the FSAR which is automatically updated each year.

Duke Power successfully deleted these figures from the McGuire Tech Specs recently on this ba sis.

However, since VEGP is an NTOL, and this is really a generic issue, we were not allowed to make the deletion.

58. Page 6-6:

Our proposed revision to 6.2.3.3 was not acceptable to the staff.

59. Page 6-7:

Specifications 6.4.1.2 and 6.4.1.3 were revised as noted in our November 6, 1986 submittal (GN-1166).

60.

Page 6-8, item 1: We were asked to provide additional. justification for our proposed revision.

(This was provided in our November 14, 1986 submittal (GN-1184).)

61.

Pages 6-9, 6-10, 6-13 and 6-14: These pages were revised as noted in our November 6,1986 submittal (GN-1166).

62.

Page 3/4 4-29: We agreed to put in the bases a reference to the FSAR table which will contain the specimen withdrawal schedule.

(This was pro rided in our November 6,1986 submittal (GN-1166).)

63.

Page B 3/4 4-10: We will provide a more legible figure.

(This was provided in our November 6,1986 submittal (GN-1166).)

64.

Page 3/4 3-43: The staff questioned the fact that we did not include all of our seismic instrumentation in the tech specs. They also questioned the fact that we did not require an analog channel operational test on the triaxial response spectrum analyzer. We responded to the first concern via a telecon subsequent to this meeting by reviewing R.G.1.12 against our proposed tech specs with the reviewer.' He agreed with our proposed list of instrumentation with the exception that we should include the instrument at auxiliary building floor elevation 220. We agreed and provided a marked-up table in our November 6, 1986 submittal (GN-1166).

The second question was resolved by explaining to the reviewer that the response spectrum analyzer was not an on-line device at VEGP. The operator h. - to physically remove recorded data f rom the seismic instrumental.on and insert the tapes into the analyzer for processing.

The ref ore, the analog channel operational test is not applicable.

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Pese 10 65.

Page 3/4 3-42:

The staff will have to give our proposed revision further consideration.

66.

Page 3/4 6-9:

The reviewer will have to review our justifications for I

proposed revisions to items a and d.

67.

Page 3/4 4-21:

The staff will have to give our proposed revision further consideration.

68.

Pages 3/4 7-19, 7-20 and B 3/4 7-5:

The staff informed us that the STS i'

were in error by correlating snubber failures on a system by systes basis.

They have found that the majority of snubber failures cannot be related to the system on which they are installed.

Therefore, the final draf t of the VEGP Unit 1 tech specs will be revised accordingly.

69.

Page 3/4 4-15:

faulty steam generator tubes was unacceptable in the absence of an I

analysis providing a plugging limit for the sleeved tubes.

The 40%

plugging limit is not applicable to sleeved tubes.

70.

Page 3/4 4-17:

4 Our proposed revision was deemed appropriate.

]

71.

Page 3/4 4-18:

According to the staff the Code of Federal Regulations states that steam generator tubes which fall into Category C-3 constitute a condition which requires prompt notification.

revision to this table was unacceptable.

Therefore our proposed 72.

Page 3/4 4-14:

The addition of the words " Condition IV" was unacceptable i

to the staff.

i i

73.

Page 3/4 4-36:

The staff requested that we clarify the wording found in 4.4.10.2.

Our November 6,1986 submittal reflects the revised wording.

74.

Page 3/4 6-2: The staff stated that ANSI N 45.4-1972 is required by the Code of Federal Regulations. However, in our FSAR we are committed to 4

ANSI N 56.8-1981. GPC will have to check this in order to ensure i

compliance with the law.

75.

Page 3/4 6-6:

The staff questioned our limits on containment pressure and we referred to sections 6.2.1.1.1 and 6.2.1.1.3 of the SER which indicates the acceptability of these limits.

76.

Page 3/4 8-1:

This page was revised as noted in our November 6,1986 submittal.

Footnote # was added to address a concern expressed by the

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reviewer arising from a TDI program requirement.

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77.

Page 3/4 8-3:

We agreed to clarify the FSAR concerning diesel-generator j

start timess i

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Page 11 78.

Page 3/4 8-4:

This page was revised as noted in our November 6,1986 submittal.

One of the consenters from the regional staff called to our attention a letter to J. Nelson Grace concerning some additional fuel oil testing for inorganic zine contamination. During the ensuring discussion it became clear that tech specs were not the only issue since NRR made the statement that additional testing may not resolve the question of inorganic zine contamination.

This issue will have to be resolved in the realm of the FSAR and any questions concerning tech specs will have to be deferred until such resolution is achieved.

79.

Page 3/4 8-6:

Proposed revision were found to be acceptable.

At this point in the meeting we returned to the ICSB reviewers concerns.

80, On page 3/4 3-51 under Instrument 14: Expand entries to include wide range and narrow range instruments following the plants' R.G.1.97 Type A designation for these instruments as discussed in the staff's SER.

Response: This entry was revised appropriately.

81.

In Table 3.3-8:

Add appropriate entries for containment isolation valve positions following the plant's R.G. 1.97 categorization of this instrumentation as Category 1.

Response

The VEGP FSAR states that containment isolation valve position indication is a Category 2 parameter.

The position indication meets the intent of Category 1 requirements with the exception of redundant indication.

We subsequently agreed to add this parameter to the table in the tech specs.

82.

On page 3/4 3-18 under Functional Unit 5:

Provide appropriate entries for feedwater isolation on low Tavg coincident with P-4 asshown in Figure 7.2.1-1 (Sheet 13) of the FSAR.

Response: The feedwater isolation on low Tavg coincident with reactor trip is not assumed in any of the Vogtle FSAR Chapter 15 non-LOCA safety analyses. It is considered to be a diverse means of isolating main feedwater.

It should be noted that this function was assumed in the Vogtle superheat steamline breaks (outside containment) mass / energy analysis to conservatively provide the earliest feedwater isolation signal available during the transient. However, because this modeling assumption is used to increase the severity of the transient, it is not considered necessary for reactor protection and is not considered necessary for inclusion in the Vogtle tech specs.

The staff, however, insisted that the final draf t include this functional unit.

- Attachment 1-Page 12

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

Response: The Proof and Review Copy was corrected appropriately, t

84. On page 3/4 3-30 under " TABLE NOTATIONS:" Verify correctness of "_"

for T2 in note *.

Response: The Proof and Review Copy was corrected appropriately.

85. On page 3/4 5-2 under Section 4.5.1.2:

Required surveillance deviates from the Westinghouse STS.

Provide justification.

Response: The final draft will include the requirement to perform an analog channel operational test on the accumulator level and pressure channels once per 31 days. We argued that it is more appropriate to verify accumulator level and pressure by checking the actual values using control board instrumentation rather than relying on the alarms. The reviewer agreed with our position but stated that our proposed revision was generic and could not be granted for VEGP at this time.

86.

In Tables 2.2-1 and 3.3-3:

Correct entries under " TRIP SETPOINT" and

" ALLOWABLE VALUE" to agree with those provided in WCAP-11269,

" Westinghouse Setpoint Methodology For Protection Systems Vogtle Station."

Response: This was resolved by our October 13, 1986 submittal (GN-1115).

87.

On page B 3/4 3-3: Correct the discussion provided under "P-11" to agree with that provided in Table 7.3.1-3 of the FSAR.

Response: The bases were reworded as noted in our November 6 submittal.

88. 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) of the FSAR.

Entry under " ACTION" deviates from Westinghouse STS.

Provide justification.

89. On page 3/4 3-16 under Functional Unit 3.a.4:

Entry under " ACTION" deviates from Westinghouse STS.

Provide justification.

Response: This item was resolved by discussion at the meeting. No revision of the Proof and Review Copy was required.

90. On pages 3/4 3-17 and 3/4 3-18 under Functional Unit 4: Reference to footnote f deviates from Westinghouse STS.

Provide justification.

Response: This item was resolved by discussion at the meeting.

l

Pase 13 91.

In Tables 3.3-2, 3.3-3, and 4.3-2: Add instrumentation for all other ESFAS functions,'such as control building ESF electrical equipment rooms HVAC, that are not included in other sections of the technical specifications.

Response: This item was resolved by discussion at the meeting.

92. On page 3/4 3-19 under Functional Unit 6:

Prov.de appropriate entries for auxiliary feedwater pump suction transfer following the guidance of the Westinghouse STS.

Response: VEGP is not equipped with this feature.

93. Controlled Leakage: At this point in the meeting, we participated in a conference call between Westinghouse and the NRC staff during which Westinghouse explained the justification for our proposed revision to this specification. The staff asked about the intent of this specification and Westinghouse responded that the intent was to ensure proper ECCS flow by limiting flow to the seals. The intent is not to detect excessive seal leakage at the reactor coolant pumps. The staff was also concerned about maintaining seal injection throttle valve position. These valves are enclosed within protective covers which are locked under administrative control. Thus positive protection is afforded to prevent inadvertent movement of the valves. At this point the reviewers had no further questions and agreed that our proposed revision would be acceptable.

94 Pages 3/4 3-42 and 3-44: The staff will have to give our proposed revision further consideration.

95.

Page 3/4 6-9: This page will require further consideration by the staff.

96.

Page 3/4 0-2: The staff would not accept our proposed revision to this specification. GPC stated that they wished to pursue this item further with the staff.

97.

Page 3/4 4-21: The staff could not accept our proposed revision since the following criteria are not met except for the RHR suction isolation

valves, 1.

Full closure of pressure isolation valves indicated in control room.

2.

Interlocks to prevent opening when RCS pressure is above low pressure system design pressure.

3.

High pressure alarms.

98.

Page 3/4 8-11: This page was revised as noted in our November 6,1986 submittal in response to a comment from the regional staff.

Pase 14 99.

Page 3/4 8-12:

Our proposed revisions were accepted. We will have to revise the FSAR to be consistent with the tech specs, 100.

Page 3/4 8-13:

We need to verify that the allowable value for float voltage for each connected cell (2.10 volts) will be sufficient to maintain the battery in the fully charged condition.

(We provided this in our November 14, 1986 submittal - GN-1184.)

101. Pages 3/4 8-15 and 8-16:

We agreed to reinstate the STS version of footnote

• in response to concerns expressed by the reviewer.

102.

Page 3/4 8-18:

The reviewers comments were resolved by discussion at the meeting.

103.

Pages 3/4 8-19 and 8-20:

The reviewers comments were resolved as noted in our November 6,1986 submittal. We agreed to add a reference to the table of overcurrent protective devices to the bases.

This was provided in our November 6,1986 submittal.

104.

Page 3/4 8-21:

The reviewer was not willing to allow Table 3.8-2 to be deleted from the tech specs and removed to the FSAR.

We agreed to replace the table in the tech specs. We also resolved the reviewers concern about the deletion of the surveillance requiring calibration of the thermal overload protection devices.

105.

Page 3/4 3-19:

The reviewer accepted our proposal to replace Action 19 with Action 23 as applied to Functional Unit 6e.

Also, our proposed action 28 applied to Functional Unit 6f was acceptable.

106.

Page 3/4 6-4:

This page was revised as noted in our November 6,1986 submittal.

107.

Page 3/4 6-11:

This page was revised as noted in our November 6,1986 submittal.

108.

Page 3/4 6-15:

Our proposed revision was found to be acceptable.

109.

Page 3/4 6-16:

This page was revised as noted in our November 6, 1986 submittal and a reference to the containment isolation valve table was added to the bases.

110.

Page 3/4 7-9:

This page was revised as noted in'our November 6,1986 submittal.

The allowed outage times will have to appear as "laters" until the staff reviews our PRA justifying our proposed allowed outage times.

111.

Page 3/4 7-13: This specification was revised as noted in our November 6, 1986 submittal.

112.

Page 3/4 4-37:

We will have to address the two control valves in the head vent flow paths. Are these valves required to function in order for the head vents to be operable?

Attachment _i_,,

Pase 15 113. Page 3/4 3-70: The final draf t will contain surveillance requirements similar to those which appear in the Hope Creek Tech Specs rather than a reference to our Turbine Overspeed Protection Reliability Progr.sm.

We will discuss this further with the staff.

114. Page 3/4 8-4: The staff stated that the frequency at which accumulated water is checked for and removed from the diesel fuel oil tank, iaould be 31 days inctead of 92 days regardless of where the fuel oil :anks are in relation to the water table. This is part of the McGuire diesel fuel oil surveillance package.

115. Page 6-15: Our proposed revisions were accepted.

116. Page 3/4 6-14: Our proposed revision was accepted but we have to provide the eductor flow rates as soon as possible.

117. Page 3/4 3-21: Action 18 was revised as noted in our November 6,1986 submittal.

118. Page 3/4 6-1: The reviewer questioned our revision concerning the blind flange on the fuel transfer canal. They will consider this further.

119. Page 3/4 6-15: We have to provide additional information concerning the equivalence of the heat removal capacity of the containment spray system and the containment fan coolers. Are two trains of spray equivalent to the fan coolers? If not, we will have to revise the action statements appropriately.

The following paragraphs contain the resolution of the remaining concerns of the ICSB reviewer.

120. One page4 3/4 7-12 under Section 4.7.4 br Since the nuclear service cooling water system pumps start is initiated automatically by more signals than 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 instrumentation under Tables 3.3-2, 3.3-3, and 4.3-2.

On page 3/4 7-11 under Section 4.7.9.br Since the component cooling water system pumps start is initiated automatically by more signals than just safety injec. tion 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 instrumentation under Tables 3.3-2, 3.3-3, and 4.3-2.

Response: The above concerns are addressed by response time testing and the loss of offsite power test performed pursuant to diesel generator surveillance requirements.

I Attachment l_ _,

Page 16 121. In Table 4.3-2: Entries under " SLAVE RELAY TEST" should be changed to "M" in lieu of "Q" per Section 7.3.3.3 of the staff's SER. Also, the addttional relief provided by Enclosure 3 to the staff's July 24, 1985 letter related to WCAP-10271 should be considered.

Response

We addressed the ites pertaining to slave relay testing in our August 27, 1986 submittal (GN 1064). We feel that we have taken full advantage of the relief provided by Enclosure 3 to the staff's July 24, 1985 letter related to WCAP 10271.

122. In Table 3.3-4:

Cross-references to other tables should be corrected.

Response: We resolved this in our October 13, 1986 submittal (GN-1115).

123. On page 3/4 3-49:

Delete "[Illustrational Only}" and verify that all appropriate entries have been included in these tables.

(

Responsc The staff accepted our proposed revision to this specification. They did state however that they felt that auxiliary feedwater flow and steam generator pressure should be added to the table of remote shutdown instrumentation. We need to justify our position that our proposed table contains the necessary instrumentation.

124. On page 3/4 3-68 under " ACTION":

Clarify "less than required" if this is intended to refer to " Minimum Channels Operable."

Response

This was resolved as noted in our November 6,1986 submittal.

125. 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 instrumentation for those other signals are not now included under Table 4.3-2, include appropriate surveillance requirements or, a stated above, include the instrumentation under Tables 3.3-2, 3.3-3, and 4.3-2.

Response

This ites was resolved by the fact that we are revising the FSAR such that the electrical penetration room exhaust air cleanup system will no longer be referred to as an "ESF" system. Therefore since no credit is taken for the operation of this system, it will not appear in the tech specs.

i 126. On pages 3/4 7-15 and 7-16:

Entries under Section 4.7.6.e.2 and Section 4.7.6.a.5 appear to be duplicate. Verify that separate entries are required.

Response

This was resolved by discussion at the meeting.

i l

Pese 17 127. On page 3/4 7-29 under Section 4.7.11.b and c Since the ESF roon

. cooler systes is initiated by more signals than 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 l

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

Response: This ites was resolved by discussion at the meeting.

128. In Table 3.3-11 under " Electric Steam Boiler Isolation":

Correct the conflict between the entries under " Instrument channel" and Figure 7.6.6-6 of the FSAR.

l Response : The FSAR will be revised to resolve the conflict.

l 129. On page B 3/4 3-1:

Include the second paragraph a reference to the staff's February 21, 1985, SER on WCAP-10271. Also in the fourth line from the bottom insert a "+" in the equation.

Response

The appropriate revisions were made to the bases.

l 130. On page 3/4 3-12 under " TABLE NOTATIONS": Note 11 should be made l

consistent with corresponding note 12 of Generic Letter 85-09 to req 2 ire l

independent testing of the undervoltage and shunt trip attachments of the reactor trip breakers for each train every 62 days on s' staggered l

test basis.

i Response: The final draf t will be revised so the STS are consistent i

with Generic Letter 85-09.

131. In Section 6.7: Add progranaatic requirements for surveillance and i

controls for restoration of inoperable instruments for all plant l

post-accident monitoring instrumentation that are classified Category 2 or 3 per R.C.1.97 following the guidance provided in an October 12, 1983, meno from Roger J. Mattson to Darrell G. Eisenhut covering technical specifications for post-accident monitoring instrumentation.

Response

We will have to discuss this further with the staff. The final draft will be revised to include these requirements.

132. Specification 4.0.3: the staff will consider deleting the sentence from the bases which states that action statements re entered when the surveillance requirements should have been performed rather than at the time of discovery.

133. Page 3/4 3-41:

Proposed revision was found to be acceptable.

1 t

Pam 18 134 Pages 3/4 5-1, 5-4 ar.d 5-6:

Proposed revisions were found to be acceptable.

135. Page 3/4 9-1:

Proposed ' revisions were found to be acceptable.

136. Specification 3.7.1.1: Our proposed revision was not acceptable in the absence of a safety crsluation by Westinghouse.

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MEMORAfFJUM FOR: Thomas M. Novak, Assistant Director Division of Licensing FROM:

James P. Knight, Acting Director Division of Engineering

SUBJECT:

REQUEST TO AMEND SUSQUEHANNA UNIT 1 TECHNICAL SPECIFICATIONS (TAC # 55139)

References:

1. Letter from N.11. Curtis to A. Schwencer, " Proposed Arrendment 43 to License No. NPF-14," dated May 18, 1984

2. NRC Generic Letter 84-13 " Technical Specification for Snubbers," dated May 3,1984 In Reference 1, the licensee proposed changes to Appendix A of License No.

1 NPF-14, the Susquehanna Unit 1 Technical Specifications. One of these proposed changes involves Technical Scecification Section 3/4.7.4, j

" Snubbers." The licensee is eliminating Table 3.7.4-1, " Safety Relcted Mechanical Snubbers" which is a tabular listing of all safety related mechanical snubbers in Unit 1.

The elimination of the tabular listing of snubbers in the Technical Specifications agrees with policy established in a recent generic letter (Reference 2) and is acceptable.

In addition to the above change, the licensee proposed a revision to the Unit 1 Technical Specifications which would place the inspection schedules of snubbers l

on a system basis rather than a plant or unit basis. This should be denied because it would be in conflict with established NRC policy and precedent and also in conflict with an industry standard (OM-4) which is nearing final acceptance for issuance. The title of the standard is " Examination and Performance Testing of Nuclear Power Plant Dynamic Restraints (Snubbers)."

You should be aware that the issue of " plant versus system" as the basis for inspection schedules was thoroughly debated within the CM-4 Working Group on Dynamic Restraints and it was agreed that the intent of the standard was to W rvey all the snubbers in a M snt to locate those which may suffer from an

< dentified failure mode which is not system specific.

If the failure mode can be narrowed to a specific design, model or type of snubber then only those snubbers must be inspected. However, this selection must be made from the entire unit or plant population of snubbers, not just from the system in which the valid failure mode was identified.

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'We recognize that this denial will cause Units 1 and 2 to be different. We only recently became aware of the fact that'Susquehanna Unit 2 and several other plants were issued technical specifications in conflict with our policy and that of the industry standard. We have evaluated the effect of such operation and have concluded that it does not present -a hazard to the health and safety of the public to operate in this manner in'the time period required to rectify this matter.

This concludes the MEB action for TAC # 55139.

James P. Knight, Acting Director Division of Engineering cc:

D. Crutchfield, DL R. Bosnak, DE A. Schwencer, DL M. Campagnone, DL E. Butcher, DL H. Bramer, DE ll

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Vo6744 Justification for Changes to Specification 4.7,4'of then?-e-d Ge'f Technical specification

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Operating exaerience has indicated that snubber inoperabilities are not generally re5ated to the systems on which they are installed. Such inoperabilities are usually caused by either:

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1. an isolated incident such as installation error,

2. a problem related to the snubber design, or

3. a general environment problem, such as high temperature or radiation.

None of the above causes are system unique.

PWA A The fkch:r.ical Engineering Branch) believes that the visual inspections should be used to identify the tfpe of inoperable snubbers, and rein-spection intervals should be based on the number of failures within the identified type instead of the specific system to which the original inoperable snubber was attached.

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June 4,lh86

1. M United States Nuclear Regulatory Comission File: X780102 Juite 2900 Log:

GN-928 101 Narietta Street, N.W.

Atlanta, GA 30323 Re'ference: Vogtle Electric Generating Plant - Unit 1 Readiness Review Program Module 138, Coatings Attention:

Mr. J. Nelson Grace Your letter of March 9,1986 asks for a discussion of Georgia Power Company plans to prevent the detrimental processes described in NRC IE circular No.

77-15, which can occur when fuel oils are stored in contact with zinc.

Bechtel Engineering has recomended the purchase of fuel oil with a neutralization number of 0.1 or less, and monitoring of the fuel in storage to identify any detrimental processes. This recomendation is discussed in the Module 138. Independent Design Review Report, Section 7.4.5.3.

We have reviewed the Bechtel recomendations and, in addition, have obtained the recomendation of the coating manufacturer, Ameron. Ameron recommends that stored oil in contact with their inorganic zine coating be kept within the following specifications:

Neutralization number No more than 0.4 Mercaptan..

No more than 0.01%

The Ameron guidelines will be incorporated into procedure 30080, " Diesel Fuel Chemistry Control", in a revision to be issued by June 30, 1986. The procedure calls for sampling each storage tank quarterly and sending the samples to a qualified offsite laboratory for analysis, Laboratory analysis results will be reviewed to verify that the oil in storage meets the recomended neutralization numbers and mercaptan levels.

If the results of the laboratory analysis indicate unacceptable conditions exist, corrective actions could include, but are not limited to, jhMA b )

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1.

Pumping the oil to the start-up boiler storage tank for burning or into tanker trucks for disposal.

2.

Applying i'n oil additive to reduce the neutralization number.

3.

Adding more oil to dilute the mercaptan value.

We believe the above described program is adequate to prevent any degradation of engine performance.

Tb3s response contains no proprietary information and may be placed in the NRC public document room.

Very truly your

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c auw Docket No.: 50-261 MEMORANDUM FOR:

B. Requa, Pn) ject Manager, Project Directorate #2 Division of PWR Licensing-A FROM:

Faust Rosa, Chief Electrical, Instrumentation and Control Systems Branch Division of PWR Lic'ensing-A

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION, TECHNICAL SPECIFICATION CHANGES, GENERIC LETTER 85-09 AND ITEM 4.3 T.S. (MPA-890) 0F GENERIC LETTER 83-28 Plant Name:

Robinson 2 Utility:

Carolina Power and Light Company ce si tatus:

MC bN Resp. Directorate: PD #2/DPA Project Manager:

B. Requa Review Branch:

EICSB/DPA Review Status:

Incomplete

Reference:

Carolina Power and Light Company letter, S. R. Zimmerman to Steven A. Yarga,

"--- Response to Generic Letter 85-09",

December 20, 1985.

Generic Letter 85-09 presented the NRC staff positions with respect to Technical Specification changes (Item 4.3 T.S., MPA B-90) related to the Reactor Trip Breaker design modifications of Item 4.3 of Generic Letter 83-28.

We understand that the Carolina Power and Light Company has not submitted these proposed Technical Specifi-cation changes as yet for staff review.

However, the company by letter dated December 20, 1985, reconnended against periodic testing of the bypass breakers, based on a probablistic risk analysis by the Westinghouse Owners Group.

request for additional infonnat The enclosed testing of the bypass breakers.Qn (RAI) reaffinns the staff's position regarding The RAI also provides detailed additional guidance as to what the staff requires for review.

copy of Generic Letter 85-09, to t' he licensee.Please transnit the RAI, along with a

/5/

Faust Rosa, Chief Electrical, Instrumentation and Control Systems Branch Division of PWR Licensing-A

Enclosure:

As stated N

V.f cc:

T. Novak Distribution:

L. Rubenstein Docket File 50-26I J. E. Knight EICSB Rdg.

F. Rosa A. Toalston (PF)(2)

C. E. Rossi Of gh _LQ) }

D. Lasher Robinson 2 S/F EICSB/DPA EICSB/DPA SL/EICSB/ PA ATpa,lstyi:dd DL ey JEKnight

/b BC/EIqSB/DP3 FRosaff#g" 7/2%/86 7/ af /86 7/ Jo /86 7/ 3 e /86 0FFICIAL RECORD COPY

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REQUEST FOR ADDITIONAL INFORMATION IN ACCORDAhCE WITH GENERIC LtTTER 85-09 AND ITEM 4.3 T.S. (MPA B-90) 0F GENERIC LETTER 83-28.

TECHNICAL SPECIFICATION CHANGES H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT 2 DOCKET NO. 50-261

(

INTRODUCTION Generic Letter 85-09 contains staff positions regarding proposed Technical Specification changes pertaining to reactor trip breaker (RTB) design changes (Item 4.3 of Generic Letter 83-28).

Generic Letter 85-09 concluded that Technical Specification changes should be proposed by licensees to explicitly require independent testing of the under-voltage 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.

Staff Positions You are requested to submit proposed Technical 5pecification changes, noting particularly the following posi ions of Generic Letter 85-09.

For Table 3.3-1:

1.

For the reactor trip breakers (RTBs) and for the automatic trip logic, different actions are specified for Modes 1 and 2 as compared to Modes 3*, 4*, and 5* (asterisk means with the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal).

~

Action 12 applies to Modes I and 2, whereas Action 13 applies to Modes 3*, 4*, and 5*.

r

.. 2.

A different action is permitted (Action 14) if only one of the diverse trip features (i.e., shunt or undervoltage trip) is inoperable.

For Table 4.3-1:

4 1.

The manual reactor trip tests at refueling intervals specify testing o the bypass breaker trip circuits as well as the RT8 trip circuits Independent testing of the undervoltage (UV) and shunt trip attachmen (STA) of each RTB and bypass breaker is specified (Action 11).

2.

Each RTB is to be tested at least every 62 days on a staggered basis (Action 7).

Independent testing of the UV and STA is specified (Action 12) as part of this RTB test.

3.

Each automatic trip logic train is to be tested at least every 62 days on a staggered basis (Action 7).

4.

The reactor trip bypass breakers are to be manually tested prior to placing the breaker in service (Action 13).

Also an automatic UV trip test is to be made durkng each refueling outage (Action 14).

Some ifcensees or applicants have declined to provide Technical Specification changes for testing of the bypass breaker prior to each time it is put into service.

The reason cited is that a study by Westinghouse for the Westinghouse Owners Group (WOG) has shown only marginal reliability improveme The NRC staff finds this basis for not testing to be unacceptable.

When a reactor trip breaker (RTB) is tested, the reactor trip bypass breaker is put

,, into service as a backup to the remaining RTB. The Generic Letter 85-09 Tech-nical Specifications specify a manual trip test of the bypass breaker prior to putting it into service.

This testing'is a simple procedure and it is prudent to do this test before relying on this breaker as a. backup to the remaining RT Although the probabilistic risk analyses have shown that the testing of the bypass breeker only marginally improves overall reliability, no claim is made that the reliability is decreased or that the bypass breaker is not safety related.

Accordingly, the NRC staff concludes that this test should be made as specified in Generic Letter 85-09.

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3 0 JUL iM5 Docket No.: 50-286 MEMORANDUM FOR:

D. Neighbors, Project Manager, Project Directorate #3 Division of PWR Licensing-A FROM:

Faust Rosa, Chief Electrical Instrurpentation and Control Systems Branch Division of PWR Licensing-A

SUBJECT:

RE0 VEST FOR ADDITIONAL INFORMATJON, TECHNICAL SPECIFICATION CHANGES, GENERIC LETTER 85-09 AND ITEM 4.3X.'S. (tiPA-890).;

0F GENERIC LETTER 83-28 j

Plant Name:

Indian Point 3 Utility:

New York Power Authority Docket No.:

50-286 Licensing Status:

OR Resp. Directorate: PD #3/DPA Project Manager:

D. Neighbors Review Branch:

EICSB/DPA Review Status:

Incomplete Generic Letter 85-09 presented the NRC sta'ff. positions with respect to Technical Specification changes (Item 4.3 T.S., MPA B-90) related to the Reactor Trip Breaker design modifications of Item 4.3 of Generic Letter 83-28. We understand that the New York Power Authority has not submitted these proposed Technical Specification changes as yet for staff review. The enclosed request for information (RAI) provides detailed additional guidance as to what the staff requires for review.

Please transmit the RAI, along with a copy of Generic Letter 85-09, to the licensee.

Faust Rosa, Chief Electrical, Instrumentatfor and Control Systems Branch,

Division of PWR Licensing-A

Enclosure:

/

As stated

/

cc:

T. Novak Distribution:

O S. Varga Docket File 50-286 F. Rosa 2

EICSB Rdg.

C. E. Rossi k

Contact:

A.Toalston(PF)(2)

Indian Point 3 S/F A. Toalston, EICSB/DPA D. Lasher X27243 J. E. Knight o.

EICSB/DPA EICSB/DPA SL/EICSB/DPA BC/EICJ DPA AToalston:dd DLasher JEKnight y t FRosa y l' d 3 76C 7/ 2.Y/86 7/ 9 7 /86 7/se /86 7/ J o /86 OFFICIAL RECORD COPY hbbh.

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REQUEST FOR ADDITIONAL INFORMATION IN ACCORDANCE WITH GENERIC LETTER 85-09 AND ITEM 4.3 T.S. (MPA B-90) 0F GENERIC LtTTER 83-28.

PROPOSED TECHNICAL SPECIFICATION CHANGES.

INDIAN POINT STATION, UNIT 3 DOCKET NO. 50-286 i

Introduction

,s '

Generic Letter 85-09 contains staff positions regaEding propose'd Technical g

Specification changes pertaining to reactor trip breaker (RTB) design ' changes (Item 4.3 of Generic Letter 83-28).

Generic Letter 85-09 concluded that Technical Specification changes should be proposed by licensees to explicitly require independent testing of the under-voltage 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.

Staff Positions You are requested to submit proposed Technical Specification changes, noting particularly the following positions of Generic Letter 85-09.

.For Table 3.3-1:

i 1.

For the reactor trip breakers (RTBs) and for the automatic trip logic, differe.nt actions are specified for Modes 1 and 2 as compared to Modes 3*, 4*, and 5* (asterisk means with the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal).

Action 12 applies to Modes 1 and ?, whereas Action 13 applies to Podes 3*, 4*, and 5*.

)

e i

2 2.

A different action is pennitted (Action 14) if only one of the diverse trip features (i.e., shunt or' undervoltage trip is inoperable.)

/

For Table 4.3-1:

1.

The manual reactor trip tests at refueling intervals specify testing of the bypass breaker trip circuits as well as the RTB trip circuits.

Independent testing of the undervoltage (UV) and shunt trip attachment (STA) of each RTB and bypass breaker is specified (Action 11).

2.

Each RTB is to be tested at least every 62 days on a staggered basis (Action 7).

Independent testing of the UV and STA is specified (Action 12) as part of this RTB test.

3.

Each automatic trip logic train is to be tested at least every 62 days on a staggered basis (Action 7).

4.

The reactor trip bypass breakers are to be manually tested prior to placing the breaker inkservice (Action 13). Also an automatic UV trip test is to be made during each refueling outage (Action,14).

Some licensees or applicants have declined to provide Technical Specification changes for testing of the bypass breaker prior to each time it is put into service. The reason cited is that a study by Westinghouse for the Westinghouse Owners Group (WOG) has shown only marginal reliability improvement by such testing.

The NRC staff finds this basis for not testing to be unacceptable. When a reactor trip breaker (RTB) is tested, the reactor trip bypass breaker is put O e

. into service as a backup to the remaining RTB.

The Generic Letter 85-09 Tech-nical Specifications specify a manual trip test of' the bypass breaker prior to putting it into service.

This testing is a simple ' procedure and'~lt is prudent /~

to do this test before relying on this breaker as a backup to the remaining'khB Although the probabilistic risk analyses have shown that the testing of the bypass breaker only marginally improves overall reliability, no claim is made that the reliability is decreased or that the bypass breaker is not safety related.

Accordingly, the NRC staff concludes that this test should be made as specified in Generic Letter 85-09.

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NAY 0 v ISS6

.p NEMORANDUM FOR:

M. Slosson, Project Manager, PWR Project Directorate #3 Division of PWR Licensing-A A

FROM:

C. E. Rossi, Assistant D'irector Division of PWR Licensing-A '

SUBJECT:

INDIAN POINT 2 - SUPPLEMENT TO SAFETY EVALUATION R ON REVIEW OF DESIGN FOR AUTOMA1'IC SHUNT TRIP FOR TRIP BREAKERS Plant Name:

Indian Point, Unit 2

~

Utility:

Consolidated Edison Co.,,

Docket No.:

50~247 TAC Nos.:

53176 (4.31; 55358 (MPA B-90)

Licensino Status:

OR Resp. Pro.f. Dir. :

Pn 43/0PA D oject Manager:

M. Slosson r

Review Rranch:

EICSB Review Status:

Incomplete By letters dated November 4,1983. March 16,1984, April 2,1984 and June 2 1984, the Consolidated Edison Company of New York, Inc. provided responses Generic Letter 83-28 including Item 4.3, Reactor Trip System Reliability l

(autcmatic actuation of the Shunt Trip Attachment for W plants).

Evaluation Report (SERI by the NRC staff dated June A Safety 2271984, found that the the preimplementation review requirements of Item 4.3 The SER noted however that the seismic qualification of the shunt trip attach-ment was being conducted by the Westinghouse Owners Group (WOG) and that the Itcensee should confinn that the shunt trip is seismically qualified when the results of the WOG qualifications are completed.

following implementation of the shunt trip modificationsThe SER also noted that submit proposed technical specifications which are respon,sive to the staffthe licens requirements noted in the SER.

Subsequently, Generic Letter 85-09 was issued providing guidance for the Technical Specifications associated with the shunt coil trip attachments.

The Itcensee responded by letter dated February 14, 1986 to the seismic quali-k fication and Technical Specification issues.

For the seismic qualification issue, we find the licensee's response acceptable.

For the Technical 9

Specification issue, we do not find the licensee's response acceptable...

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2-responsive to the Generic Letter 8S-09 requirem a are not forthcoming we recommend the issuance of a show cause order If this is Enclosed also is the SALP input for the addition 6 These issues 83-28, items 4.3 and 4.3 (T.S.).

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Enclosure:

As stated Charles E. Rossi, Assistant Director Division of PWR Licensing-A cc:

T. Novak B. J. Youngblood Distribution:

Document Control 016 C. E. Rossi

Contact:

EICSB Rdg.

IP-2 S/F A. Toalston, EICSB/DPA A. Toalston (PF)(2)

X27243 D. Lasher J. E. Knight F. Rosa

• SEE PREVIOUS CONCURRENCE EICSB/DPA EICSB/DPA SL/EICSR/DPA BC/EICSB/DPA AD/0PA AToalston:ct* DLasher*

JEKnight*

FRosa*

CERossi 3/

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/86 3/

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s SUPPLEMENTAL SAFETY EVALUATION REPORT DOCKET NO. 50-247 INDIAN POINT UNIT 2 GENERIC LETTER 83-28. ITEM 4.3 REACTOR TRIP BREAKER AUTOMATIC SHUNT TPIP i

INTRODUCTION AND

SUMMARY

Generic Letter 83-28 was issued by NRC on July 8[ 1983, indicating actions to be taken by licensees Eas~e~d on the generic impitcation of the Salem ATWS events.

Item 4.3 of the generic letter requires that modifications be made to improve the_ reliability of the reactor trip system by implementation of an automatic actuation of the shunt trip attachment on the reactor trip breakers.

By letters dated November 4,1983 March 15,1984, April 2,1984 and June 22, 1984, the licensee, Consolidated Edison of New York, Inc., provided responses to the plant specific questions identified by the staff in its August 10, 1983, Safety Evaluation Report on the generic proposed Westinghouse design. Subsequent to the review of the licensee's submittals, the staff issued a safety evaluation report on June 22, 1984 indicating the acceptable and unacceptable aspects of the licensee's responses and requested the licensee to further respond. By letter dated February 14, 1986, the licensee submitted further responses re-garding the Technical Specifications and seismic qualification of the automatic shunt trip. We find the response to the seismic qualification issue to be ac-ceptable, but are unable to accept the responses to the Technical Specification issues.

EVALUATION The staff identified the following concerns in its safety evaluati,on report (SER)ofJune 22, 1984 and our evaluation of each is presented below.

OA A D 1 ARbY O (Q i Gv i ( w g

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I a)

The seismic qualification of. the shunt trip attachment is being conducted by the Westinghouse Owners Group (WOG).

~

The licensee should confirm that the shunt trip is seismically qualified when the results of the WOG qualification program are completed By letter of Febrtiary 14, 1986, the licensee advised us that the seismic quali fication of the reactor trip breaker shunt trip attachment (model 08 50 been completed by Westinghouse, that the Indian Point 2 plant spe parameters were enveloped by the Westinghouse test results, and that 10 50.49 does not apply because the equipment is located in a mild env We find this acceptable.

b)

Following the implementation of the shunt trip modifications, the licensee should submit proposed technical specifications which are responsive to the staff requirements noted in the enclosed SER.

Subsequently, the staff issued Generic Letter 85-09 providing guidance the Technical Specifications.

The licensee has declined to implement this guidance in the following respects 1/

1.

The bypass breaker testing would not be included in the Technical Specifications.

J,/. Letter from John D. O'Toole to Mr. Steven A. Varga, February 14, 1986.

\\

4

. 2.

The remaining surveillance test requirements would be administra-tively controlled on an interim basis without' Technical Speciff-cation changes until the requirements can be optimally determined.

The above approach is unacceptable to the NRC staff. When a reactor trip breaker is tested, the reactor trip bypass breaker is put into service as a backup. The Generic Letter 85-09 Technical Specifications Guidance requires a manual test of the bypass breaker prior to putting it into service.

This test-ing is a simple procedure and it is prudent to do this test before relying on this breaker as a backup to the remaining reactor trip breaker.

The licensee states that probabilistic risk analyses (PRAs) show that testing of the break-er has an insignificant effect on the overall reliability.

No claim is made that the reliability is decreased or that the bypas's breaker is not safety related.

Therefore, we believe that this test should be made as specified in Generic Letter 85-09.

Secondly, the licensee does not plan to amend its Technical Specifications until the requirements can be optimally detemined.

No schedule is given for such a determination. This is unacceptable to the NRC staff.

The licensee should amend its Technical Specifications consistent with the guidance and re-quirements of Generic Letters 83-28 and 85-09.

If further changes are then required in the interest of optiaizing safety, the staff will consider further amendment requests.

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CONCLUSION Based on the above review, we find th'at the seismic qualification issue been satisfactorily resolved.

The licensee's position with respect to the Technical Specification issue is not acceptable.

The licensee should submit proposed Technical Specifications consistent with Generic Letter 85-0 for staff review.

4 6

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v Docket No.: 50-369/370 NAY g21986 MEMORANDUM FOR:

Darl Hood, Project Manager, Project Directorate #4 Division of PWR Licensing-A FROM:

Charles E. Rossi, Assistant Director for Technical Support Division of PWR Licensing-A

SUBJECT:

EICSB (PWR-A) INPUT FOR MCGUIRE UNITS 1 AND 2 TECHNICAL SPECIFICATION CHANGES, MPA B-90 Plant Name:

McGuire Nuclear Station, Units 1 and 2 Utility:

Duke Power Company Docket Nos.:

50-369, 50-370 TAC Nos.:

55361, 55362 Licensing Status:

OR Resp. Directorate: PD4/DPA Project Manager:

Darl Hood Review Branch:

EICSB/DPA Review Status:

Incomplete By letter dated December 7, 1985, Duke Power Company submitted proposed technical specification changes for the McGuire Nuclear Station, Units I and 2, in response to Generic Letter 85-09.

EICSB finds the proposed changes acceptable except for the following:

1.

The licensee should provide a proposed technical specification for testing the Reactor Trip Bypass Breaker prior,to each time it is put into service consistent with Generic Letter 85-09.

2.

The licensee should provide proposed technical specifications for verifying the operability of the Bypass Breaker Trip circuit (s) each refueling outage consistent with Generic Letter 85-09. is the SALP input for this review.

Cri;2 '.:I;.d 4, Charles E. Rossi, Assistant Director for Technical Support Division of PWR Licensing-A

Enclosure:

Distribution:

As stated Docket File 50-369/370 C. E. Rossi EICSB Rdq.

V. Benaroya cc:

T. Novak A. Toalston (PF)(2)

McGuire S/F B. J. Youngblood D. Lasher J. E. Knight

Contact:

F. Rosa A. Toalston, EICSB/DPA 69 AbM3

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SAFETY EVALUATION REPORT TECHNICAL SPECIFICATION CHANGE (MPA B-90) i MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 INTRODUCTION By letter dated December 7,1985, Duke' Power Company proposed changes to the McGuire Technical Specifications pertaining to Reactor Trip System Instrumentation and Surveillance 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 Duke Power Company's proposed Technical Specification changes are consistent with those specified in Generic Letter 85-09, except as indicated below by items 1 and 2.

1.

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

No provision is ' included for verifying the operability of the Bypass 2.

Breaker trip circuits each refueling outage.

The licensee has declined to provide Technical Speciffiations for the above items on the basis that there is no significant benefit to be expected from inclusion of Bypass Trip Breaker testing 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 as a backup.

The Generic Letter 85-09 Technical Specifications specify a manual test of the i

2 bypass breaker prior to putting it into service. This testing is a simple procedure and it is prudent to do this test before relying on this breaker as a backup to the remaining reactor trip' breaker. Although the licensee states that probabilistic risk analyses (PRAs) show that testing of the breaker has an insignificant affect on the overall reliability, no claim is made that the reliability is decreased or that the bypass breaker is not safety related.

Therefore, we believe that these tests should be made as specified in Generic Letter 85-09.

We find the remaining Techncial Specification changes proposed by Duke Power Company to be consistent with Generic Letter 85-09 and therefore acceptable.

CONCLUSION We find the Technical Specification changes propnsed by Duke Power Company for the McGuire Nuclear Station, Units 1 and 2 to be acceptable except that the licensee has declined to provide proposed Technical Specifications to provide for testing of the Reactor Trip Bypass Breakers and associated circuitry.

The licensee should submit for staff review proposed Technical Specifications for this purpose as indicated in Generic Letter 85-09.

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Docket Nos.: 50-327/328 MEMORANDUM FOR:

C. Stahle, Project Manager, Project Directorate #4 Division of PWR Licensing-A

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FROM:

C. E. Rossi, Assistant Director Division of PWR Libensing-A

SUBJECT:

EICSB(PWR-A)INPUTFORSEQUOhAHNUCLEARPLANT GENERIC LETTER 83-28 ITEM 4.3,'- T.S.(MPA 8-90)

TECHNICAL SPECIFICATION CHANGES Plant Name:

Sequoyah Nuclear Plant, Units 1 and 2 Utility:

Tennessee Valley Authority Docket Nos.:

50-327/328 TAC Nos.:

55377/55378 Licensing Status:

OR Resp. Directorate: PD #4/DPA Project Manager:

C. Stahle Review Branch:

EICSB/DPA Review Status:

Incomplete 1

By letter dated May 23, 1986, TVA submitted proposed Technical Specification changes for the Sequoyah Nuclear Station, Units 1 and 2 in response to Generic Letters 83-28 and 85-09. Other proposed Technical Specification changes are also included. The enclosed Safety Evaluation Report pertains only to the Generic Letter 83-28 and 85-09 ftems. The other changes are to be reviewed under separate TAC numbers.

EICSB finds the proposed Technical Specification changes pertaining to Generic Letters 83-28 and 85-09 acceptable except for the following:

j,,

1.

For the Reactor Trip System Instrumentation, Table 3.3-1, Modes 3*,4*,5*

should be separated from Modes 1 and 2 for Functional Units 20 (Reactor Trip Breakers) and 21 (Automatic Trip Logic), and an action statement added, similar to Action 13 of Generic Letter 85-09, for Modes 3*,4* and 5* (asterisk means with the reactor trip system breakers in the closed position, the control rod drive system capable of rod withdrawal, and fuel in the reactor vessel).

2.

Table Notation (9) of Table 4.3-1 should be expanded to test the operability 4

of the Bypass Breaker trip circuits.

Contact:

A. Toalston, EICSB/DPA t

X27243 x

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79 d ~LL h @l%

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C. Stable 3.

A Functional Unit should be added to Table 4.3-1 for Ch Tests of the Reactor Trip Bypass Breaker (including Table Notations) similar to that for Functional Unit 23 of Generic Letter 85-09. is the SALP input for this review.

Crisical o1::ned by,'

Charles E. Rossi, Assistant Director Division of PWR Licensing-A

Enclosure:

As stated cc:

T. Novak B. J. Youngblood Distribution:

Docket File Nos.: 50-327/328 EICSB Rdg.

A. Toalston (PF)(2)

D. Lasher J. E. Knight F. Rosa C. E. Rossi V. Benaroya Sequoyah S/F E!CSB/DPA EICSB/DPA SL/EICSB/DPA l

A g gon:ct DLgr hJEKnight BC/EICSB,/

F0B/DPA FRosa y g VBenaroya CER 6/J3/86 6/AC/86 6/ 2.'//86 T/ 2. /86 1/3/86 y /86 0FFICIAL RECORD COPY y

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SAFETY EVALUATION REPORT TECHNICAL 7 tIIICATION CHANGES [MpA B-90)

SEQUOYAH NUCLEAR STATION, UNIT 5 I AND 2 ~

t INTRODUCTION 1

By letter dated May 23, 1986 TVA proposed changes to the Sequoyah Technical i

Specifications pertaining to Reactor Trip System Inittrumentation j

in response to Generic Letter 85-09 (MPA B-90).

ance Generic Letter 85-09 concluded that Technical Specification changes should be proposed by licensee o explicitly require independent testing of the undervoltage and shunt trip attach j

the reactor trip breakers during power operation, testing of bypass br a ers prior to use, and independent testing of the control room manual s s

and wiring during each refueling outage.

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

t 1.

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

i (asterisk means with the reactor trip system breakers in the close tion, the control' rod drive system capable of rod withdrawal, and fue the reactor vessel) for Functional Units 20 and 21 of Table 3.

2.

No provision is included for testing of the Reactor Trip Bypass Breaker b

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.

OL

2 The proposed Technical Specification changes for Functional Units 20 (React 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*.

1 For Modes 3*,4* and 5*,

the-reactor trip breakers should be opened if the condition is not remedie n

a given time period as specified in Action 13 of Generic Letter 85-09.

TVA has declined to provide Technical Specification changes for surveillanc testing of the reactor trip bypass breakers on the basis that Westinghouse O wners Group (WOG) calculations have shown no significant reliability improvemen m

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 tri p or remote undervoltage trip) prior 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 (fttle Notation 11) also re-quires a manual trip function test to independsW ly

• irify the Operability of the undervoltage and shunt trip circuits of A bg., Breakers each refueli ng outagi.

This test is important to assure that the bypass breaker' can be manu-ally trirped from the control panel.

CONCLUSION The licensee should amend its proposed Technical Specification changes by i

)

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 Breakt.r 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.

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

9 4

l

EICS8 SALP INPtif PLANT:

'Sequoyah. Units 1 and 2 i

SUBJECT:

Proposed. Technical Specification Changes, MPA B-90 l

EVALUATI(NI PERFORMANCE CRITERIA CATEGORY EASIS l

Management N/A-Wo-basis for assessment.

~ '

I Involvement 1

App a ch to 2

A13 0 ugh not c0

".as provjded,,:,nsistent with staff position, supporting justification Resolution of.

l Technical Issues Responsiveness

.2

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

i i

2 1

Enforcement i

l History N/A No basis for assessment.

s L

^

Reportable Events,

N/A No basis for assessment.

i Staffing N/A No basis for assessment.

. "I"I"9 N/A No basis for assessment.

i

/

t

. Issue (b)

Leak Testing of Pressure Isolation Valves Request for Additional Infomation 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 continuously 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.

4 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.

Comment Technical Specification Index Page IV: Teble 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.

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 XII: 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 positiont. 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, from 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 these 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.

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

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 response times.

Pressurizer pressure-low, reactor coolant pump-under voltage, and reactor coolant pump underfrequency are only applicable above the P-7 interlock. This is not as conservative ~ as the guidance 2

i D

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 bb not'd'that there is no e

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 this table as there were in Table 3.3-2.

There are no funtional units covering i

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 1

RCP's, nor any ESFAS interlocks at P-12 or P-14.

i 3

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

-PORY position indication

-PORV Block position indication

-Safety Valve Position

-Plant Vent Monitors

-Containment Isolation Valve Position Table 3.3-10 There is no. criteria 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 "RER" 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.

4

_ ~ -.

3.4.4a The phrase,"because of excessive 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 la hours,.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 4

valve (s) and remove power from the block 4

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

O notification to the NRC pursuant to Table 4.4-2 (cont'd) specification 6.9.1'. Alsolunder 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."

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

() b '

using a one percent standard gas or(STS say use b equivalent thereof.

and 4% for both Bydrogen and Oxygen.

The limits in this table are too lenient Table 3.4-2 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.

ST rc L

Table 3.4-4 is non-conservative with 5

deble 3.4-4

, d,'/

respect to the values given in the STS.

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

Specification deleted, why?

4.4.9.1.2 The maximum auxiliary spray water temperature differential is given as 625 3.4.9.2c degrees which is nearly twice the 320 degrees in the STS.

reliance on SAFETY CONCERN-Typically, 3.4.9.3 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.

i 6

3.4.11 The site specific design has three 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.

4.4.11.1

~

There is nd quarterly surveillance on the operation of the RCS vent path block valve.

3.5.1 Delete reference to instrument span in 3.5.1'o. This will eliminate the confusion between actual volume requirements and what an instrument reads.

4.5.1.1 STS has a provision for verifying avery 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 occurs. Why this was granted to Callaway is unknown, but does not justify carte blanche for Vogtle.

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

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

i 7

J 4.6.1.7.2 See wr'ite-up.above. If

• note remains, then e'ach time "14-inch" appears, an

• should be placed,.

I 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 i

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.

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.

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

t 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.

Table 3.7.1 Under percent of rated thermal power, 65

)

should be 64 and 43 should be 42.

8 i

l i

Table 4.7.1 S.TS 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 deleted.

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 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 2njection test signal.

9

,w y_,..-

-e

-.---..---.--------,,,------.,,-,-,,-y-

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.

4.'7.6.a Remove [ ] around 80 (degrees).

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 i

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 4

still leave seven bus transfers that 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 compensate for the fact that fuel oil will be stored in contact with inorganic zine 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 i

that this test should be integrated into TS 4.8.1.2.4.

10

' s 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 is 450.

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

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

i 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 bat tery 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.

i i

Table 4.8-2 s

The float voltages and specific gravities are not appropriate. The float voltages should be:

h&

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 I

condition.

I 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.

11

.,-----,n-..-,_,

_,,--,_,.._,_.n

_,_n--.nn

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 specification 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 same 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 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 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.

12

,-,.y c

____.-__.r__,___.__.

__.,_,_m.,

-_____._,___m.

,,...,,.c.

,_._m

4.11.2.5 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 samp1s 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 and the FSAR section 1.9.9.5.

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.

13

3/4 6.4 FSAR section 1.9.7 states that conformance is to RG 1.27 Rev. 2 Jan.

1986, not RG 1.27, Har. 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".

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

-_..-,______,__._,-_,--_.____y-e-

.-.-..,..,r.,

+

these temporary racks will be removed leaving no st9 rage 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.

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

The system is designed for only one cycle at power.

Administrative Controls This item deviates from the requirement to have a corporate officer, not the 6.1.2 GMVNO perform this. See THI item I.C.3 o

and I.A.1.2.

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

6.2.2e

1) Only overtime that exceeds the guidelines is reviewed. The STS hasin

(,

\\de Y.which is the only means of monitoring Q

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

h8

2) The level of authorization is too low t

it for the approval of excess overtime, resides at the same level that approves regular overtime.

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

d Each procedure of 6.7.1 above, and changes thereto, Shall be Suggest that 6.7.2 read:

the responsible and approved by the GMVNO or the department head ofa i

h in to implementation and reviewed periodically as set fort department, d approve the administrative procedures. The GMVNO shall review an following procedures:...

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

2) The GMVNO does not review allc 15

i to read,"The change is documented, reviewed, and approved in accordance

. with section 6.7.2 within 14 days of implementation.

Clarify who accepts-suggest:"b shall 6.12.2b become effective upon review and acceptance by the PRB and approval per Specification 6.7.2

~.

Same as for 6.12.2b above.

6.13.2b General Comments

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

" LOOSE PART DETECTION SYSTEM" be in the T

2) Will TS 3.3.3.8,If it is to be deleted, then TS 3.3.3.9 and higher need 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 METHODO Technical Specification Tables 2.2-1 and 3.3-3 to eliminate 9

1986.

non-conservative values. Submitted to the NRC September 2,

4.4.1.2.2, 4.4.1.3.2 & 3.4.1.4.1b Also affects TS's

5) Several Tables have been removed from the TS which is pr information of a TS upgrade program. Proper references to where the TS, then now resides must be made. If not directly as part o revised and read as if the tables still exist.

" pursuant to Specification

6) In several action statements the words,6.9.2" appear. All the special report section.

16

.c n

t.

g#

f f*a h blii G-

st#

v!X'

?-

b. The analyses take credit for reactor trip from the power range s

high neutron flux channels. The power range high neutron g". 7 FN flux channels are not required to be operable during shutdown

q

/

by the Technical Specifications. The source range and.

e 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 t!r 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 pr: posing changos to Taoles 3.3-1 and 3.3-2 of the Technical Specifications which will a) require two source range channels to be operable in modes 3-5 whenever +.he 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

..g "l

'T 3 4., -

p %r. * *. ~

..y 'i-e.*

e 97i.,

channels. This value is consistent with the safety analysis assumptions and with f

.. ?..i

~

the existing specification for the power range channel. Note that for the case of y.

4,=~,

a rod withdrawal from subcritical, several seconds would elapse between the tiee I

when the source r ange trip sgnal is Jenerated (when the core is at or.just below critical) and the time when the protection is actually needed (significan,t power g~eneration occurring)- see, for example Figures 15.2-1 to 15.2-3 in the 'J75AR.

The 0.5 second response time, which is used for consistency with with t'.1e other channels, is therefore more than adequate to protect against rod withdrawal events in modes 3-5. We have determined that response time testing of these char nels can be implemented with minimal difficulty.

9 e

O 4

29

=

s:. cn C1 ~

~

~

IMMDEM C1MMENTS ON V0GT.E UNIT 1 PBQW)t5GilDfTJ XFMbM5U FE ;}rg,A i g M" r

v " "

DKP FECTION 7 t

/

/j In Section 7.2.2.2 of the SER, a technical specification requirement for the turbine trip on reactor trip circuitry is discussed. Provide appropri-ate technical specification.

Proposed technical specifications include changes approved by the staff's 2

review of WCAP-10271. The staff's letter of July 24,1985, and the " West-inghouse Owners Group Guidelines for Preparation of Submittels Requesting Revisions to RPS Technical Specifications" M1sted 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.

On page 2-7 under NOTE 1: Correct the conflict between the equation used G'

here and that used on page 7.2.1-5 of the FSAR.

On page 2-9 under NOTE 3: Correct the confitet between the equation used g/

here and that used en page 7.2.1-6 of the FSAR.

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

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

two sections. W # # 8 M#### ^""

Y l

l g.

i

!4

---

a-r-m--

e en,

,,,-----e.,,,mw-

,,,,e.

.w,-r,.,.,,.

ww----,--_-.,--,------

~ ~,, - - - - - -,, -

2-i G/

On page 3/4 3-2 under Functional Unit 5.b: Entry under " MINIMUM CHANNELS a or n*

e OPERABLE" has been reduced to one. Provide, justification.

o on pages 3/4 3-g and 3/4 3-10:

Functional Unit 18.d is missing. Renumber

• • **V' r entries as appropriate.

on page 3/4 3-17 under Functional Unit 3.b.1: Rewrite entry around manual or d

initiation of containment spray and isolation. Under Functional Unit 4.4.1:

Correct the conflict between the logic' used here and that shown in Figure 7.2.1-1(Sheet 8)oftheFSAR. Under Functional Unit 3.b.5: Entries under

" MINIMUM CHANNEL 5 OPERABLE" should be revised to encompass the redundancy provided by the design.

On page 3/4 3-18 under Functional Unit 5: Provide appropriate entries for g/

feedwater isolation on low Tave coincident with P-4 asshown in Figure

[.2.1-1(Sheet 13)oftheFSAR.

/

8/0 On page 3/4 3-20 under Functional Unit 10: Provide appropriate entries p

for control room isolation on high chlorine 1nput as shown in Figure 7.2.1-1 (Sheet 8) of the F,5A.k Entry under " ACTION" deviates from Wef

@ o se STS. Provide justification.,

f ggf / p.;g e

ww9 wwe l..

O

_.,__,.._m-

,--..,,_,.._-y,,

,,.m__,.

,_c.y.-_,_,

,.. b nc

. l On page 3/4 3-16 under Functional Unit 3.a.4:

o Entry under " ACTION" devi-ates from Westinghouse STS. Providejustfficatien. / * *"# # 5 ^##

On pages 3/4 3-17 and 3/4 3-18 under Functional Unit 4:

f Reference to foot-note f deviates from Westinghouse STS. Provide justification.

/ 1 9 p H n d / * ** )

p1 In Tables 3.3-2, 3.3 3, and 4.3-2:

Add instrumentation for all other ESFAS

/

functions, such as control building ESF electrical equipment rooms H

, 'that are not included in other sections of the technical specifications.

i

/pMo preu. /' M o r A t 4 t' n ' )

s o

On page 3/4 3-1g under Functional Unit 6:

Provide appropriate entries for auxiliary feedwater purnp suction transfer following the guidance of th Westinghouse STS.

IN" # #

'# ' ##O On page 3/4 3-26 under Functional Unit 4.e:

i Entries under " TRIP SETPO!NT" y

~

and " ALLOWABLE VALUt" should be positive following tkt guidance Westinghouse STS.

o On page 3/4 3-30 under "TA8LE NOTATIONS:"

Verify correctness of " t,".

for 7 in note *.

g 9

6 g..

4 1

)

_..___..__-.....___.___,.___._._____________.._..._.._.s

~

gns,s nul tc. 20 2;E 4-

/

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

"M" in lieu of "Q" per Section 7.3.3.3 of'the staff's SER. Also,the additional relief prnvided by Enclosure 3 ko the staff's July 24,1985 letter related to WCAP-10271 should be considered, bo In Table 3.3 4: Cross-references to other tables should be corrected.

(ped semd)

I h

On page 3/4 3-49: Delete "[!11ustrational Only]" and verify that all F

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 #

and narrow range instruments following the plants' R.G.1.g7 Type A designa-tion for these instruments as discussed in the staff's SER.

In Table 3.3 8: Add appropriate entries for containment isolation valve d-i positions following the plant's R.G.1.g7 categorization of this instru-mentation as Category 1.

j r

--,-_-nm.-...---,~r-

.,. - - - - - - -. - -,, _ ~, - -

-n

_w _, m

-e.---,-.e.,r----w-~.-n-s--

h G

s

Lce s 12u:

' 7dd I

l Clarify "less than required" if this On page 3/4 3-88 under " ACTION":

is intended to refer to " Minimum Channels.0perable."

Required surveillance deviates c e On page 3/4 5-2 under Section 4.5.1.2:

(/ me 6e*WM k

7Jo Provide justification.,

from the Westinghouse STS.

Since the electrical penetration On page 3/4 6-21 under Section 4.6.6.d.2:

signals

/o room exhaust air cleanup system is initiated automatically b illance requirements for the instrumenta-j

,thanjustsafetyinjectionandsurve Table 4.3-2, include tion for those other signals are not now included under l d the appropriate surveillance requirements or, as stated abov M'd ""

<1>

3.3-2, 3.3-3, and 4.3 2.

instrumentation under Tables Sface the campenent ecoling water de un page 3/4 7-11 under Section 4.7.3.b l thanjust system pumps start is initiated automatically by more sig o

tation for safety injection and surveillance requirements for the instru include appro-those other signals are not now included under Table 4.3 ld the instru-priate surveillance requirements or, as stated above, inc

/

3.3-2, 3.3-3, and 4.3-2.

mentation under Tables se Since the nuclear service cooling JS On page 3/4 7-12 under Section 4.7.4.b:

by more signals than water system pumps start is initiated automatically l

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,., 10/10/96 13:13 NO.013 007

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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 'equirements or, as stated above, include the instrumentation under Tsbles 3.3-2, 3.3-3, and 4.3-2.

th fj o Onpages3/47 hand 7-16: Entries under Section 4.7.6.e.2 and Section 4.7.6.e.5 appear 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 e*

system is initiated by more signals than just safety injection and sur-ve111ance requirements for the instrumentation for those other signals are not now included under Table 4.3-2 include appropriate surveillance re-quirements or, as statad above, include the instrumentation under Tables 3.3-2. 3.3-3, and 4.3 2.

i' In Table 3.311 under " Electric Steam Boiler Isolation": Correct the con.

r <-

flict between the entries under "Instrumen+. Channel" and Figure 7.6.2-6 of the FSAR.

(

yj 0 On page 8 3/4 3-1:

Include in the second paragraph a reference to the v

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

e

10ns ss 13:14 NO.013 008 l /

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

.c e '

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

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1

~

..e In Section 6.7: Add programatic requirements for surveillance and centrols NJ for restoration of inoperable instruments for all plant post-accident mon-itoring instnamentation that are classified Category 2 or 3 per R.G.1.97 l

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.

. g In Tables 2.2-1 and 3.3-3: Correct entries under " TRIP SETPOINT" and

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

g In WCAP-1126g errors are included for Veritrak pressure transmitters used

(

in several protection channels.

In light of Westinghouse's recent notiff-cation to licensees concerning possibla 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.

'( s on page 3/4 3-12 under " TABLE NOTATIONS": Note 11 should be made consistent r0 with corresponding note 12 of Generic Letter 86-09 to require ' independent testing of the undervoltage and shunt trip attachments of the

• reactor trip g..

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

4

2D, P J o taa. =

w l

Vo G TLE T S 's Qng 1.

Ungineered Safety Features Actuation System Instroentatten. Table 3.3 2 fpage3/43-15) i I

htenatic Safety InTectian__(Thi h containct eressurgLis act reevived to I

k inoerania inwM 4 ty Tab

. 42. In a ie L T from J. sailey arc to

t. Denton NRC December 9 1g856 en analysis of Earts break.LOCA in Mode 4 esas provided to the staff assumine famedista and automatic actuation of i

31 at the end of blowdown. Provide revisions to either.the Safety Analysis or the Technical Specifications so that they am consistent. If you choose to revise the Safety Analysis the operator response time to annually actuate 81 should be Justifled. In other reviews the staff has secepted operator response tisws of 10 minutes following a control room elare.

2.

4teactor Trip System and Engineered Safety Features Acteatten System 1 Instrumentation (page33/488) f The bases describe t$s japortance of msponse ties testing for the f

React,or Trip and shgineered Safety postures actuation functions.

l Limiting conditions for operation are not provided in the Technical Specifications for the ressonse times or for their surveillones. ple e 1

4errect this inconsistency.

3.. Reactor Trip Instrumentation. Table 3.31 ( ge3/432)lfree The p5AR Evaluction of inadvertent control nk withdrawa suberitical or low power assumes teactor trip to be initiated by a high meutron flux signal from the power range channels (Iow setting}s.

The Technicel heirications do not require thtporar rente thanne to be omrAple wwnjhe_.reager'3EllikfP.lLYEIl~lsodes 3.e.. ena3;.

I' lease T torrect this seconsistency. If peu response is that the reactor would trip from signals generated by tu source endmentueuse$ste range channels. Provide the response tiens for this instrue ntation under 1 l

3 and demonstrate that the transient analysis in the p3AR is bounding.

Reactor Coot c.t System Not Shutdown 3.4.1.3 and $1d htdown 3.4.h and 4.

3.4.1.4.2 Weges 3/4 4-3 to 3/4 4-6)

The FSAR evaluation for inadvertent control bank witMrewal from suberitical assumes that two reactor coolant yJaps are operating. The

. Technical _ Specification do act require..., reactor coolant _ pumps _to.be in l

operautoninmesa==

=1 ar a se tros roe mana with'dr~awal transient were no occur without coolant pump flow; AR.sheminimumDNBRmightbe decreased below that calculated in the 73 provide additional safety analyses of inadvertent control rod withdrawal transients in modes 4 ana 5 without reacter coolant pump flow or demonstrate that inadvertent criticality from sentrol rod bank withdrawal cannot occur in Modes 4 and 5.

NC.213 212

• 12/15 56 12:15 j

-t-j Main $ team Line Isolatten Valves J (Page 3/4 7-9) and Engineered 5.

3-17)y Features Actuetten Systems a trunentattens fa61e).3,2 fpage

'Safet t

/

The Safety Evaluatten in the PSAR for steam generator tube ruptum

(

assumes that the operator takes action to isolate the leak by closing the N31V en the associated steam line. The Technteel Senef ficet' ens do not,

require manus 1 flo1at< en maa*M11oi 5F eseracir ty o' ' * "C va <a made rrovise aceitiona' safety ans'yses of an un'sola",able steam generator re accident in Mode 4 er provide Technical Specifications l

eensisten the current Safety Analysis.

J,'

'5.

Steam Line Atmospheric Relief Valves The FSAR safety Evaluation of Steam Generator Tube aspture assumes operator action to open the atmospheric relief valves en the steam lines ef the unaffected steam generators. This action is fulvited to limit radiation release to the atmosphere if offsite power is Isst. The Technical Specifkat< aan de est yoguire operability of the, atmospheric reim wrin s.

Prov< de ad:!stiera sr.rsy unnyw. or a steam generator tube rupture accident with inoperable steen line relief valves er provide

~

Technical Specificattens that are eenststent with the Safety Analysis.

.7.

Special Test faception 3/4.10.4 Reactor Coolant Leops (page.3/4.10.4) power Operation in Mode 1 is positted to the p-7 interlock setpoint which may be in excess of 105 power. The FSAR does met evaluate power operation without the Teactor ecolant pumps. Either provide a supporting

/ Safety Evaluation er revise the Technical Specifications so that they are

'g {conistentwiththeexistingsafetyahalpls.,

~

l l

l l

-_mJ g ENTATION TURBINE OVERSPEED PROTECTION 3/4.3.4 LIMITING CONDITION FOR OPERATION At least one Turbine Overspeed Protection System shall be OPERABLE.

3.3.4 APPLICABILITY

MODES 1, 2, and 3.

ACTION:

With one stop valve or one control valve per high pressure turbine steam line inoperable and/or with one intermediate stop valve or one a.

intercept valve per lou pressure turbine steam line inoperable, restore the inoperable valve (s) to OPERABLE 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 af fected steam line(s) or is 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 />.

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.

b.

SURVEILLANCE REQUIREMENTS The provisions of Specification 4.0.4 are not applicable.

4.3.4.1 The above required Turbine Overspeed Protection System shall be 4.3.4.2 demonstrated OPERABLE:

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 a.

through at least one complete cycle from the running position:

Four high pressure turbine stop valves, 1) t _ rM x. c t.. '.. '.. c q,,

2)

P;

';f, pcc::1 :

Six low pressure turbine intermer. ate stop valves, and Z/)

3 /)

Six low pressure turbine intercept valves.

At least once per 31 days while in MODE I and while in MODE 2 wit the turbine operating, by direct observation of t b.

position,

,,y g, fu, f(,$$,, $,f,,, pgpf gQgg, 4

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

At least once per 40 months by disassembling at least one cf each the above valves and performing a visual and surface inspection o

' d.

valve seats, disks and stems and verifying no unacceptable flaws or

[(

corrosion.

hr&ist endre h (1st)ading&$ntNifgitsikrt.

Valett) 3/4 3-91 CATAWBA - UNIT 1

MEMORANDUM FOR: Darrell G. Eisenhut, Director Division of Licensing g g gg FROM:

Roger J. Mattson, Director Division of Systems Integration

S1J9 JECT:

TECHf!! CAL SPECIFICATIONS FOR POST ACCIDEf:T M0?!ITORING IftSTRUl:ENTAT!0fl e

In accordance with NRR Office Letter No. 38 the Instruwntation and Control Systems Branch (ICSR) is proposing a change to the Standard Technical Specifi-cations (STS). The change will update the STS to reflect the recornmendations contained in Regulatory Guide I.97 Revision 2, " Instr 6 mentation For Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following an Accident". As Enclosure I we are providing the supporting rationale for this change. As Enclosure 2 we are providing revised STS pages.

It is our recorrendation that the revised STS he impinented imediately on OL reviews and be included in the next publication of Westinghouse, Combustion Engineering, Babcock and Wilcox and General Electric STS.

Wafnalsvadh WJ.6%

Roger J. Mattson, Director Division of Systems Integration DISTRIBUTION:

Enclosures:

Central File As stated ICSB R/F M. Virgilio (PF)(2) cc:

C. Thomas J. Joyce C. Rcssi

Contact:

F. Rosa M. Virgilio, ICSR AD/RS Rdg.

X29454 R. Mattson T. Denning ICSB Branch Merrbers

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ENCLOSURE 1 RATIONALE FOR PROPOSED STANDARD TEChMICAL

~

SPECIFICATION CHANGE i

The Standard Technical Specifications (STS) currently impose limi h

st-conditions for operation and surveillance requi,rements on t e po t

accident monitoring instruments and include examples of the instru-i l specifica-mentation that should be included in an applicant's techn ca Using the guidance contained in the bases section of the tion submittal.

ific table STS each applicant for an operating license develops a plant spec listing the post-to be included-in the technical specification submittal The bases section of the current STS accident monitoring instrumentation.

1.97 "Instrumenta-references the December 1975 edition of Regulatory Guide Plant Conditions tion For Light-Water-Cooled Nuclear Power Plants to Assess Learned Durjng and Following an Accident" and NUREG-0578, "TMI-2 Les The proposed Task Force Status Report and Short-Term Recommendations".

l issued change contained herein updates the STS to reflect the recent y idance as to Regulatory Guide 1.97, Revision 2 and provides additional gu t's technical speci-i which instruments should be included in the appl can fication submittal.

t for all licensees of operating reactors, applican s On December 17, 1982 sent a copy operating licenses and holders of construction permits were This letter of Generic Letter No. 82-33 (Supplement I to NUREG-0737).

9 h

included guidance on post-accident monitoring instrumentatio Regulatory Guide an endorsement of Regulatory Guide 1.97, Revision 2.into 1.97, Revision 2 divides the post-accident monitoring instrume

. three categories providing a graded approach to requirements depending on the importance to safety of the measurement of a specific variable.

Cate-gory 1 includes the most stringent requirements and is intended for key variables. Category 2 includes less stringent requirements and generally applies to instrumentation designated for indicating system operating e

status and instrumentation provided to furnish information regarding the release of radioactive materials.

Category 3 is intended to provide re-quirements that will ensure high-quality, off-the-shelf instrumentation is obtained and applies to backup and diagnostic instrumentation.

Al -

though the Regulatory Guide does not include explicit guidance on tech-nical specifications it does state that the Category 1 instrumentation "should be available prior to an accident except as provided in paragraph 4.11 " Exception", as defined in IEEE Standard 279 or as specified in the Technical Specifications" (C.1.3.1).

For Category 2 instrumentation the Regulatory Guide states: "the out-of-service interval should be based on normal technical specification requirements on out-of-service for the sys-tem it serves where applicable or where specified by other requirements" (C.1.3.2).

Based on the guidance contained in Regulatory Guide 1.97, Revision 2 we recommend that the instrumentation provided to indicate system operating status, furnish information regarding the release of radioactive mater-ials and furnish diagnostic information (Category 2 and 3 as defined in Regulatory Guide 1.97 Rev. 2) be deleted from the post-accident monitor-ing instrumentation specification.

We recommend that the illustrative table provided in the STS be revised to include only key variables (Category 1) consistent with the recommendations contained in Regulatory Guide 1.97, Revision 2.

For those instruments of lesser importance

• d provided to indicate system operating status, furnish information r ing the release of radioactive materials and furnish diagnostic info Programs" tion we recommend a change to STS Section 6.8 " Procedures and d surveil-to provide the appropriate limiting conditions for operation an Through a revision to STS'Section 6.8 programmatic lance requirements.

2. and 3 requirements will provide periodic surveillance of the Category instrumentation and appropriate actions for cases when Categor These programmatic requirements include pre-instruments are inoperable.

be used planned operating procedures that address back-up instruments t i

in the event that Category 2 or 3 instruments are inoperable and ble strative controls for returning the inoperable instruments to opera These requirements provide incentives to status as soon as practicable.

l t operation.

restore instruments to operable status without restricting p an f

In addition, we recommend revising the bases section of the STS h technical ence Regulatory Guide 1.97, Revision 2 and to specify that t e ly specifications on post-accident monitoring instrumentation addre Although plant specific key variables as defined in the Regulatory Guide.

ters as considerations will necessitate having different monitored parame l

t key variables the appropriate guidance is provided for develop specific technical specifications in the revised bases section.

fies the We believe that the proposed STS revision contained herein clari instrumenta-staff's position with regard to technical specifications on following an acci-tion to assess plant and environs conditions during and dent and, therefore, reduces the uncertainty as to what the sta The impact on siders acceptable in the area of post-accident monitoring.

m

-w

e e

. industry should be positive in that the proposed STS revision narrows the scope of the instrumentation explicitly covered by limiting conditions for operation and surveillance requirements to those instruments that monitor key variables.

1 i

l 9

oo

Ei f'g ENCLOSURE 2 u,

TABLE 3.3 1

ACCIDENT MONITORING INSTRUMENTATION i

REQUIRED MINIMUM NO. OF NO. OF INSTRUMENT (ILLUSTRATIONAL ONLY)

CHANNELS CHANNELS Neutron Flux 2

1 Coolant Level in Reactor 2

1 BWR Core Thermocouples RCS Pressure 4/ core quadrant 2/ core quadrant 2

1 Drywell Pressure 2

1 Drywell Sump Level 2

1 Primary Containment Pressure 2

1 Primary Containment Isolation Valve Position 2/ Valve 1/ Valve Radioactivity Concentration in Circulating Primary Coolant 2

1 Drywell Drain Sumps Level 2

1 i

Suppression Pool Water level 2

1 Containment Hydrogen Concentration 2

1 Drywell Hydrogen Concentration 2

1 Containment Oxygen Concentration 2

1 Drywell Oxygen Concentration 2

1 Reactor Building Area Radiation 2

1

o a

TABLE 3.3 ACCIDENT HONITORING INSTRUMENTATION REQUIRED MINIMUM NO. OF NO. OF INSTRUMENT (ILLUSTRATIONAL ONLY)

CHANNELS CHANNELS

!#44eutron Flux 2

1 f#aR,CSHotLegWaterTemperature GD /,.Aar 1

/4CS Cold Leg Water Temperature C

I

!J RCS Pressure 2

1

/ 4 Coolant level In Reactor 2

1 2

V Containment Sump Water Level (Wide Range) 2 1

/

ontainment Pressure 2

% k // A 1

4 Cd9tainment Isolation Valve Positions 2/ Val ve 1/ Valve

/

re Exit Temperat'ure 4/ core quadrant 2/ core quadrant ps o E.

edioactivity Concentration In Circulating Primary Coolant 2 1

Containment Hydrogen Concentration 2

1

/-*#Ptessurizer Level

~

2M%

1

' team Generator Level (w.4-.)

d 2 steam generator 1/ steam generator

/*/

densate Storage Tank Water level 1

tainment Area Radiation (High Range) 2 1

Auxiliary Feedwater Flow (BgPlants) 2 1

e v

BASES Accident Monitoring Instrumentation The OPERABILITY of the accident monitoring instrumentation ensures that A

sufficient information is available on selected-plant parameters to monitor and assess these variables following an accidenti The instrumentation in-cluded in this specification are those instruments provided to monitor key variables, designated as Category 1 instruments following the guidance for classification contained in Regulatory Guide 1.97 Revision 2, "Instrumenta-tion For Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident."

d I

h 9

ALL STS 3-i

~ ~ ~

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6.8 PROCEDURES AND PROGRAMS 6.8.4 The following programs shall be established, implemented and maintained:

A program which will ensure the capabiljty to monitor plant variables and systems operating status during and following an accident. This program shall include those instruments provided to indicate system operating status and furnish information regarding the release of radioactive materials (Category 2 and 3 instrumentation as defined in Regulatory Guide 1.97 Revision 2) and provide the following:

(1) preventise maintenance and periodic surveillance of instrumentation.

(ii) pre-planned operating procedures and back-up instru-mentation to be used if one or more monitoring instruments become inoperable.

(iii) administrative procedures for returning inoperable instruments to OPERABLE status as soon as practicable.

ALL STS,.,,,,--ye,,

--