Text

Responds to 850503 Request for Comments on Ucs 850418 Concern That Matl False Statments Made by Clark (Gpu Nuclear) at Commission 850418 Meeting Re Development of Facility Procedures to Handle Steam Generator Tube Ruptures
	ML20128K757
Person / Time
Site:	Three Mile Island
Issue date:	05/17/1985
From:	Cunningham G; NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD)
To:	Malsch M; NRC OFFICE OF THE GENERAL COUNSEL (OGC)
Shared Package
ML20128K764	List: ML20128K757; ML20128M331; ML20128M337;
References
	NUDOCS 8505310216
	Download: ML20128K757 (9)
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4 MAY 171985 i

MEMORANDUM FOR: Martin G. Malsch '

Deputy General Counsel FROM: Guy H. Cunningham, III Executive Legal Director

SUBJECT:

UCS LETTERS ON TMI-1 RESTART This memorandum responds to your May 8th request for coments on letters dated April 18 and May 3,1985, sent to the Comission by the Union of Concerned Scientists (UCS). Pursuant to your conversation with Jack Golcberg on May 15, 1985 information concerning the May 3rd letter, which is being prepared by the Office of Nuclear Reactor Regulation and Region I, will be provided to you by close of business on May 21, 1985. The remainder of this memorandum provides comments regarding the April 18th letter which were developed in consultation with the NRR staff.

With respect to the April 18th letter, you were particularly interested in the question as to whether, in light of the information provided in UCS' letter and the remarks made by Mr. Clark of GPU Nuclear at the April 18th Comission meeting, Mr. Clark made material false statements to the Comission. Mr. Clark's remarks before the Comission in response to the Comissioners' questions would qualify as " statements" under the Commission's interpretation of Section 186 of the Atomic Energy Act. See Pucific Gas & Electric Co. (Diablo Canyon Nucicar Power Plant, Units 1 & 2),

CL1-82-1, 15 hRC 225 (1F87). 8 It is not at all apparent, however, that Mr. Clark's statements were false. The statements at issue are Mr. Clark's responses to similar questions ~

posed by different Comissioners during the course of his presentation at the April 18th Comission meeting on the TMI-1 steam generators. The Com- ,

missioners' questions were aimed at probing whether, as UCS alleges, changes were made in plant operating procedures in order to account for the degraded condition of the TMI-1 steam generators. Mr. Clark denied that the changes were proposed for that reason, explaining, for example in response to Comissioner Zech:

We believe that the procedures we now have approved for TMI-1 are the proper procedures to use for the plant. That they provide greater protection of the public health and safety, and that they should be And adopted regardless of steam generator condition.

in fact in many cases -- and I hesitate to say "all" because we have not gone back and looked -- we were f

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MAY 17 1985 looking at those changes in the procedures before the steam generator problem was even known....

Theyarenotdevelopedinordertoaccomodate[ sic) any concern about the steam generator. They are developed for other good and sufficient reasons independent of that.

Transcript at 93; see also Transcript at 68-69, 72, 78-79, 101, 108-109.

UCS points to a statement made in a GPU Nuclear Technical Data Report (TDR) as an apparent contradiction of Mr. Clark's statements:

Since extensive circumferential cracking was discovered in approximately 1200 of the 31,000 tubes, it became clear that a revised set of procedures for dealing with both single and multipleSGTRs[steamgeneratortuberuptures]

should be developed.

TDR No. 406 at 14 (Rev. 3,Dec. 2, 1983).

From one perspective, this one sentence from the TDR is not wholly consistent with the statements made by Mr. Clark at the Commission meeting. However, the sentence could also be read as merely saying that GPU should be prepared for the possibility of a tube rupture accident because one could happen. The TDR as a whole focuses on ways to improve procedures to handle tube rupture accidents; it does not promote reliance on such procedures to " accommodate" or " compensate for" degradation of the steam generators.

From the staff's viewpoint, the TMI-1 steam generators have been returned to their original licensing basis and meet applicable requirements, and therefore the improved procedures are not necessary to account for any i alleged degradation of the steam generators. The staff believes that the TMI-1 procedures are prudent procedures. In this regard, it is worth noting that, regardless of whether there is a history of steam generator tube -

degradation, a steam generator tube rupture is a design basis accident for which the staff expects a licensee to be prepared to respond. NRC and industry have had under consideration improvements in responding to such accidents so as to minimize their consequences and new Babcock & Wilcox STGR guidelines for other plants are similar to those at THI-l in many respects. The staff's interest in the emergency procedures for handling a tube rupture at TMI-1 was based primarily on the results of its review of the 1982 tube rupture event at the Ginna plant. See NUREG-0909 and NUREG-0916. In fact, as a result of knowledge gained by the Ginna event the staff asked GPU Nuclear to review its procedures for dealing with a tube rupture accident. GPU Nuclear indicated that it was considering revisions to the procedure (Mr. Clark indicated at the Commission meeting that some changes were under consideration as early ss 1979).

As described in the staff's safety evaluation, NUREG-1019, the staff considers k

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-G- , MAY 17 1985 TMI-1 procedures an improvement over the earlier procedures. Nonetheless 'the staff could accept the earlier procedures as adequate.

Thus, notwithstanding the single anomalous statement in the TDR, it is difficult to conclude that Mr. Clark's statements to the Commission were falseormisleading.*] While our review has not been an exhaustive one, the available evidence would not appear to warrant a deeper review.

Original signed by Guy H Cunnin;; ham,til Guy H. Cunningham, III Executive Legal Director

/ Mr. Clark acknowledged the statement in the TDR in a letter dated April 26,1985, which responds to UCS' letter. Although Mr. Clark does not attempt to harmonize the statement in the TDR with his statements, he reiterates his belief that the new procedures are '

in themselves meritorious. One could argue that Mr. Clark omitted to tell the Commission about the statement in the TDR or at least explain it, but it is hard to see now such an omission could be an

" obvious" or " central" one that can be reached under section 186.

See Virginia Electric & Power Co. (North Anna Power Station, Units 1 &

YT CLI-76-22, 4 NRC 480, 489 (1976).

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MAY 171985 MEMORANDUM FOR: Martin G. Malsch '

Deputy General Counsel FROM: Guy H. Cunningham, III Executive Legal Director

SUBJECT:

UCS LETTERS ON THI-1 RESTART i

This memorandum responds to your May 8th request for coments on letters dated April 18 and May 3,1985, sent to the Comission by the Union of Concerned Scientists (UCS). Pursuant to your conversation with Jack Goldberg on May 15, 1985 information concerning the May 3rd letter, which is being prepared by the Office of Nuclear Reactor Regulation and Region I, will be provided to you by close of business on May 21, 1985. The remainder of this memorandum provides comments regarding the April 18th letter which were developed in consultation with the NRR staff.

With respect to the April 18th letter, you were particularly interested in the question as to whether, in light of the information provided in UCS' letter and the remarks made by Mr. Clark of GPU Nuclear at the April 18th Commission meeting, Mr. Clark made material false statements to the Comission. Mr. Clark's recarks before the Comission in response to the Commissioners' questions would qualify as " statements" under the Commission's interpretation of Section 186 of the Atomic Energy Act. See Pscific Gas & Electric Co. (Diablo Canyon Nuclear Power Plant, Units 1 & 2),

CL1-E2-1, 15 hRC 225 (1982).

It is not at all apparent, however, that Mr. Clark's statements were false. The statements at issue are Mr. Clark's responses to similar questions ~

posed by different Comrt toners during the course of his presentation at the April 16th Comission . *ing on the TMI-1 steam generators. The Com-missioners' questions were aimed at probing whether, as UCS alleges, changes were made in plant operating procedures in order to account for the degraded condition of the TMI-1 steam generators. Mr. Clark denied that the changes were proposed for that reason, explaining, for example in response to l

Comissioner Zech:

We believe that the procedures we now have approved for TMI-1 are the proper procedures to use for the plant. That they provide greater protection of the public health and safety, and that they should be adopted regardless of steam generator condition. And in fact in many cases -- and I hesitate to say "all" because we have not gone back and looked -- we were

MAY 17 1985

', looking at those changes in the procedures before the steam generator problem was even known....

They are not developed in order to accomodate [ sic]

any concern about They are developed for othe,the steam r good and generator.

sufficient reasons i

independent of that.

Transcript at 93; see also Transcript at 68-69, 7?, 78-79, 101, 108-109.

UCS points to a statement made in a GPU Nuclear Technical Data Report (TDR) as an apparent contradiction of Mr. Clark's statements:

Since extensive circumferential cracking was discovered in approximately 1200 of the 31,000 tubes, it became clear that'a revised set of procedures for dealing with both single and multipleSGTRs[steamgeneratortuberuptures]

should be developed.

TDR No. 406 at 14 (Rev. 3,~ Dec. 2,1983).

From one perspective, this one sentence from the TDR is not wholly consistent with the statements made by Mr. Clark at the Commission meeting. However, the sentence could also be read as merely saying that GPU should be prepared for the possibility of a tube rupture accident because one could happen. The TDR as a whole focuses on ways to improve procedures to handle tube rupture accidents; it does not promote reliance on such procedures to " accommodate" or " compensate for" degradation of the steam generators.

From the staff's viewpoint, the TMI-1 steam generators have been returned to their original licensing basis and meet applicable requirements, and therefore the improved procedures are not necessary to account for any alleged degradation of the steam generators. The staff believes that the TMI-1 procedures are prudent procedures. In this regard, it is worth noting that, regardless of whether there is a history of steam generator tube -

degradation, a steam generator tube rupture is a design basis accident for which the staff expects a licensee to be prepared to respond. NRC and industry have had under consideration improvements in responding to such accidents so as to minimize their consequences and new Babcock & Wilcox STGR guidelines for other plants are similar to those at TMI-l in many respects. The staff's (

interest in the emergedcy procedures for handling a tube rupture at TMI-I was based primarily on the results of its review of the 1982 tube rupture event .

i at the Ginna plant. . See NUREG-0909 and NUREG-0916. In fact, as a result of knowledge gained by the Ginna event the staff asked GPU Nuclear to review its ;

t procedures for dealing with a tube rupture accident. GPU Nuclear indicated that it was considering revisions to the procedure (Mr. Clark indicated at the i Comission meeting that some changes were under consideration as early as 1979).

As described in the staff's safety evaluation, NUREG-1019, the staff considers g

i

-G- , gg 17 1985 TMI 1 procedures an improvement over the earlier procedures. Nonetheless 'the staff could accept the earlier procedures as adequate.

Thus, notwithstanding the single anomalous statement in the TDR it is difficult to conclude that Mr. Clark's statements to the Commission were false or misleading.*/ While our review has not been an exhaustive one, the available evidence would not appear to warrant a deeper review.

Original signed by Guy H. Cunnin; ham,111 Guy H. Cunningham, III Executive Legal Director

--*/ Mr. Clark acknowledged the statement in the TDR in a letter dated April 26, 1985, which responds to UCS' letter. Although Mr. Clark does not attempt to harmonize the statement in the TDR with his statements, he reiterates his belief that the new procedures are in themselves meritorious. One could argue that Mr. Clark omitted to tell the Commission about the statement in the TDR or at least explain it, but it is hard to see now such an omissinn could be an

" obvious" or " central" one that can be reached under section 186.

See Virginia Electric & Fower Co. (North Anna Power Station, Units 1 &

E CLI-76-22, 4 NRC 480, 489 (1976).

DISTRIBUTION ELD Rdr WDircks NRC Central HDenton ;

Subject (PA.85-26) HThomoson ['

Reading HSilver Burns chron CMcCracken Cunningham chron EChristenbury 'l Murray info JGray i

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MAY 17 1985 MEMORANDUti FOR: Martin G. Malsch '

Deputy General Counsel FROM: Guy H. Cunningham, III Executive Legal Director

SUBJECT:

UCS LETTERS ON TMI-1 RESTART This memorandum responds to your May 8th request for comments on letters dated April 18 and May 3,1985, sent to the Comission by the Union of Concerned Scientists (UCS). Pursuant to your conversation with Jack Goldberg on May 15, 1985 information concerning the May 3rd letter, which is being prepared by the Office of Nuclear Reactor Regulation and Region I, will be provided to you by close of business on May 21, 1985. The remainder of this memorandum provides comments regarding the April 18th letter which were developed in consultation with the NRR staff.

With respect to the April 18th letter, you were particularly interested in the question as to whether, in light of the infonnation provided in UCS' letter and the remarks made by Mr. Clark of GPU Nuclear at the April 18th Comission meeting, Mr. Clark made material false statements to the Comission. Mr. Clark's remarks before the Comission in response to the Comissioners' questions would qualify as " statements" under the Commission's interpretation of Section 186 of the Atomic Energy Act. See Pucific Gas & Electric Co. (Diablo G nyon Nucicar Power Plant, Units 1 & 2),

CL1-82-1, 15 fiRC 225 (1982).

It is not at all apparent, however, that Mr. Clark's statements were false. The statements at issue are Mr. Clark's responses to similar questions~

posed by different Comissioners during the course of his presentation at the April 10th Comission meeting on the THI-1 steam generators. The Com-missioners' questions were aimed at probing whether, as UCS alleges, changes were made in plant operating procedures in order to account for the degraded condition of the TMI-1 steam generators. Mr. Clark denied that the changes were proposed for that reason, explaining, for example in response to Comissioner Zech:

l We believe that the procedures we now have approved for TMI-1 are the proper procedures to use for the plant. That they provide greater protection of the public health and safety, and that they should be adopted regardless of steam generator condition. And in fact in many cases -- and I hesitate to say "all" because we have not gone back and looked -- we were l

l

MAY 171985 looking at those changes in the procedures before the steam generator problem was even known....

Theyarenotdevelopedinordertoaccomodate[ sic) any concern about the steam generator. They are developed for other good and sufficient reasons independent of that.

Transcript at 93; see also Transcript at 68-69, 72, 78-19, 101, 108-109.

UCS points to a statement made in a GPU Nuclear Technical Data Report (TDR) as an apparent contradiction of Mr. Clark's statements:

Since extensive circumferential cracking was discovered in approximately 1200 of the 31,000 tubes, it became clear that a revised set of procedures for dealing with both single and multiple SGTRs [ steam generator tube ruptures]

should be developed.

TDR No. 406 at 14 (Rev. 3,' Dec. 2, 1983).

From one perspective, this one sentence from the TDR is not wholly consistent with the statements made by Mr. Clark at the Commission meeting. However, the sentence could also be read as merely saying that GPU should be prepared for the possibility of a tube rupture accident because one could happen. The TDR as a whole focuses on ways to improve procedures to handle tube rupture accidents; it does not promote reliance on such procedures to " accommodate" or " compensate for" degradation of the steam generators.

From the staff's viewpoint, the TMI-1 steam generators have been returned to their original licensing basis and meet applicable requirements, and therefore the improved procedures are not necessary to account for any alleged degradation of the steam generators. The staff believes that the TMI-1 procedures are prudent procedures. In this regard, it is worth noting that, regardless of whether there is a history of steam generator tube -

degradation, a steam generator tube rupture is a design basis accident for which the staff expects a licensee to be prepared to respond. NRC and industry have had under consideration improvements in responding to such accidents so as

, to minimize their consequences and new Babcock & Wilcox STGR guidelines for other plants are similar to those at TMI-l in many respects. The staff's interest in the emergency procedures for handling a tube rupture at TMI-1 was based primarily on the results of its review of the 1982 tube rupture event at the Ginna plant. See NUREG-0909 and NUREG-0916. In fact, as a result of knowledge gained by the Ginna event the staff asked GPU Nuclear to review its procedures for dealing with a tube rupture accident. GPU Nuclear indicated that it was considering revisions to the procedure (Mr. Clark indicated at the Comission meeting that some changes were under consideration as early as 1979).

As described in the staff's safety evaluation, NUREG-1019, the staff considers

1

-G- . MAY 17 1985 ;

TMI-1 procedures an improvement over the earlier procedures. Nonetheless, the staff could accept the earlier procedures as adequate.

Thus, notwithstanding the single anomalous statement in the TDR, it is difficult to conclude that Mr. Clark's statements to the Commission were false or misleading.*/ While our review has not been an exhaustive one, the available evidence would not appear to warrant a deeper review.

Original signed by Guy H. Cunnin; ham Ill Guy H. Cunningha,m, III Executive Legal Director

-*/ Mr. Clark acknowledged the statement in the TDR in a letter dated April 26,1985, which responds to UCS' letter. Although Mr. Clark does not attempt to harmonize the statement in the TDR with his statements, he reiterates his belief that the new procedures are in themselves meritorious. One could argue that Mr. Clark omitted to tell the Commission about the statement in the TDR or at least explain it, but it is hard to see now such an omission could be an

" obvious" or " central" one that can be reached under section 186.

See Virginia Electric & Power Co. (North Anna Power Station, Units 1 &

YJ CLI-76-22, 4 NRC 480, 489 (1976).

DISTRIBUTION ELD Rdr WDircks NRC Central HDenton Subject (MA-85-26) HThompson .

Reading HSilver Burns chron CMcCracken Cunningham chron EChristenbury Murray info JGray Lieberman info JGoldberg RH/KC/LC info TMurley C :0EL : DEL

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UNION OF i CONCERNED ;

SCIENTISTS me ce.. mi .. we.... s.w. . s. mi . wa.hi., tem. oc 20036. <202> 2,6.seoo i DOC MEitt, USHRC l

May 3, 1985 T5 MAY-6 All :52 ,

Nunzio J. Palladino, Chairnan Thoc as M. Roberts, Comissioner GFFICr er c:h/[jy James K. Asselstine, Cocznissioner 00Chf7]sG'f '

sRANCH Frederick M. Bernthal, Coccissioner Lando W. Zech, Cocrnissioner U. S. Nuclear Regulatory Comission

Washington, D. C. 20555 Gentlecien:

On April 17, 1985, the NRC staff briefed the Comission on the status of Three Mile Island Unit 1 (TMI-1) in preparation for a restart vote. UCS's ccernents the following day on the staff's presentation were limited to the subject of the steam generators. We are writing now to address other subjects because the staff's presentation was not complete.

Specifically, the staff's presentation failed to inform you of the following:

1. Sec:e " lessons learned" items (NUREG-0737 items) previously certified by the staff to be complete are, due to new infomation, no longer complete;

2. Sccie conditions of the TMI-1 plant do not conform with the conditions

described in sworn testimony during the restart proceeding which were relied upon by the Licensing Board and Appeal Board in deciding contested issues in favor of restart; and

3. Many of the new TMI-1 technical specifications p oposed by GPU and !

approved by the staff effectively reverse Licensing Board and Appeal Board findings that certain equipment is necessary to protect public health and safety by permitting operation of IMI-1 with that equipment inoperable.

Specific examples of each of these safety deficiencies are described in detail in the attachment to this letter. A brief sumary of those examples follows.

1.a.' A lessons learned requirement was that the pressurizer safety valves be tested to demonstrate acceptable perfor=ance under accident conditions.

After tests were brought to the Appeal Board's attention showing that ine.or-rect valve adjustments caused valve damage and inadequate relief capacity, GPU l ccanitted to specific valve settings that had been shown to produce acceptable '

valve perfomance. The issue was largely resolved on this basis. However, GPU later changed the settings, as discoverd by an NRC inspector. Documenta-tion that purportedly supports the new settings is dated in early 1984 but was not scheduled to be submitted by GPU until March 31, 1985. UCS has not received that' information or the staff's evaluation of it.

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ic 26 Church i t . Cambridge. Massachusetts 02238 . (617) S47 5552

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1.b. We MI-2 accident demonstrated that additional instrumentation to detect inadequate core cooling was required. The incore thermocouples are part of that instrumentation and a lessons learned requirement was that the incore thermocouple instrumentation must be environmentally qualified. GPU 1 infomed the staff that environmental qualification has not been established, !

testing is underway, efforts to procure replacement equipment are being pur- '

sued concurrently, and if replacement is necessary, procurement and installa-tion will take 40 weeks. On March 29, 1985, the staff approved GPU's request for extension of the March 31, 1985 environmental qualification deadline on l the basis that the hot leg temperature detectors (which provide an input '

signal to the subcooling margin meters) are an adequate substitute for the I incore themocouples. However, it was apparently overlooked that the hot leg temperature instruments can not be used unless the reactor coolant pumps are running, which they will not be in certain accident conditions. In addi' tion,- '

the TMI-1 technical specifications permit plant operation with the subcooling margin instrumentation inoperable on the basis that the incere thermocouples are an adequate substitute.

2. Another lesson learned requirement was that provisions to supply emergency electrical power to the presurizer heaters must be made without degrading the capacity, capability or reliablility of the onsite emergency power supply. The boards-found the 'IMI-1 provisions to be acceptable in large part on the basis that two conditions would be satisfied: a) the heaters would not be connected until the load on the emergency power supply was stabilized; and b) that the undervoltage trip used to (.isconnect the heaters to protect

_ the power supply from a failure in the heater circuits would be set at a specified voltage and time delay. GPU has since changed the setpoints for both the voltage and time delay on the basis that starting of motors on the emergency power supply would result in tripping th'e heaters. This indicates that the heaters will be connected without first stabilizing the load on the power supply. The staff it asisaaware of item."

this situation but has taken no action

]

l other than to classify " closed l

! 3. A principal purpose of the restart proceeding was to decide, in view of the IMI-2 accident, what new safety equipment was necessary and sufficient to protect the health and safety of the public during operation of TMI-1.

That process has been completely frustrated by the approval of technical specifications that permit operation of TMI-1 with equipment inoperable which the boards found was necessary to protect the public. For example, the following new safety equipment falls in this category: subcooling margin instrumentation, position indication for the pressurizer ?ORY and safety valves, emergency feedwater system flow instruments, and reactor coolant system high point vents.

UCS believes that the Comission should take the following actions:

1. Direct the staff to identify any additional " lessons learned" require-ments or other items that the staff has certified to the Commssion as complete but which are no longer complete.

2. Direct the staff to identify any additional instances where the p present status of TMI-1 does not conform with the GPU commitment and/or board-imposed ' license condition that was relied. upon in the adjudicatory decisions favorable to restart.

/

I .

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

For items 1. and 2. above, the staff should also be directed to propose a method and schedule for resolving each safety issue identified, giving due consideration to the rights of the other parties. ~

3. Direct the staff to review the mI-1 technical specifications and amend them as necessary to require, as " limiting conditions for operation,"

the operability of all equipment relied upon by the boards as a basis for decision that WI-1 can be operated without endangering the health and safety of the public. In addition, the staff should be o'rected to serve the parties with a complete set of the amended technical specifications and solicit the parties' coments on whether the amended technical specifications contain all such required limiting conditions for operation.

In sun, UCS believes that there are more items requiring resolution p' rior to a Cor:rnission vote on restart than were identified by the staff in its April 17, 1985 presentation to the Cocrnission. We anticipate the charge that these questions are being raised at the last minute. On the contrary, these are not new questions. 'Ihey are issues which were purportedly settled during the restart proceedings but which have become unraveled by either inadvertent or intentional lack of attention to the conditions of restart imposed by or implicit in the ASLB and Appeal Board decisions.

Sincerely, Robert D. Pollard Nuclear Safety Engineer N .

)

Ellyn . Weiss General Counsel encl: Additional Safety Issues Requiring -

Resolution Prior To M I-1 Restart cc: WI-1 Service List 9

ADDITICNAL SAFETY ISSUES RECUIRING RESOLUTION PRIOR TO THI-T RESTART Union of Concerned Scientists DOCKETED usnRc May 3, 1985 ,

85 MAY -6 A11 :52

1. " Lessons Learned" Recuirements That Are Incceplete 5'? W 07EECDET:.7Y

a. NUREG-0737. Item II.D.1, PerformanceTestingofBoilihgdfais.ykhd? ::(

Pressurized-Water Reactor Relief and Safety Valves (NUREG-0578, Section 2.1.2)

One of the lessons learned requirements was that reactor coolant system safety valves had to be tested under the full range of expected operating con-ditions for design basis transients and accidents, which would include solid-water and two-phase flow conditions. [Ref. 1.] As discussed below, data from testing Dresser Industry safety valves, including the specific model used as pressurizer safety valves at INI-1, showed that certain plant modifications and adjustments 1/ to the safety valves were necessary to prevent valve chatter that could reduce relief capacity and damage the valve internals. GPU commit-

~

'ed to these changes, which were relied upon by the Appeal Board to resolve the issue. Recent information indicates that GPU will not utilize the valve settings relied on by the Eoard.

j,/ Inese adjustments are called rinc settings. Although there is some inter-action between the three separate ring settings, the principal effect of each ring setting is as follows: the lower ring . setting affects the rate at which the valve opens and adjusts " accumulation," the pressure above the p. eset opening pressure at which the valve is fully open; the middle ring adjusts " blowdown," the pressure below the opening pressure at which the valve is fully reclosed; the upper ring setting is used to compensate for the effects of backpressure downstream of the valve. Accumulation and blowdown are generally expressed as a percentage of the pressure setpoint.

For example, a safety valve set to open at 2500 psi with 3% accumulation and 10% blowdown should open at 2500 psi, be fully open at 2575 psi and reclose at 2250 psi. [Ref. 2.)

/

)

B&W provided the 'following evaluation of the results from the testing performed on the safety valves:

Recent review of the EPRI test data indicate that for certain combinations of_ lower and middle ring settings. the safety valves could chatter (rapid cycle valve opening and closing).

Two consequences of valve chattering have been noted during valve testing: a reduction in safety valve capability to relieve pressure while chattering and subsequent damage to valve internals. During the EPRI tests, this chattering phenomenon was noted only with valves with long inlet piping configurations. However, present understanding of the effect '

j of ring settings on valve operation indicates that even with .

the valves mounted directly on the pressurizer nozzles, valve chatter could be postulated for a combination of ring settings significantly different from those used during the EPRI tests.

[Ref. 3.3 During the restart hearings conducted by the Licensing Board, UCS assert-sd, inter alia, that appropriate qualification testing of the safety valves had not been performed to verify the capability of these valves to function during normal, transient, and accident conditions. [Ref. 4.] In resolving this question, the Licensing Board found that a valve testing program being conducted by the Electric Power Research Institute (EPRI) was adequate to reveal any design deficiencies in the safety valves used at TMI-1. [Ref. 5.3 This subject was revisited during the reopened hearings held by the Appeal Board in March 1983. GPU testified that as a result of the EPRI tests

"[t3he safety valves were shown to be acceptable under all modes of operation after certain modifications were made." [Ref. 6.3 The modifications which GPU testified were necessary in order to assure acceptable safety valve perform-ance were to change "the safety valve inlet piping from a long inlet to - a short inlet arrangement" [Ref. 7.3, and to readjust the safety valve ring set-tings to "the final EPRI ring settings which we used in the last five tests."

[Ref. 8.3 GPU's comr.11tment to the latter modification has been abandoned.

O

N During an NRC inspection of TMI-1 conducted on January 21-24, 1985 (but not reported until March 18 1985), the inspector noted that GPU had changed the safety valve ring settings. GPU furnishe'd the inspector with documenta-tion from EaW dated in February and May of 1984 that purportedly supports the revised ring settings. However, the inspector concluded that since the staff's safety evaluation report had been based on the earlier ring settings, there was a need to inform the ' Office of Nuclear Reactor Regulation (NRR) of the details. GPU representatives agreed "and committed to furnish the ring satting information to NRR by 3/31/85." [Ref. 9.] To date, UCS has not received that info mation or any staff evaluation of it.

UCS believes that this safety issue must be resolved before restart. The safety significance of the ring settings for the pressurizer safety valves is illustrated by the fact that after reviewing their safety valve ring settings against the EPRI test results, Duke Power Company declared the safety valves inoperable and shut down Oconee Unit 2. [Ref. 10.] The Oconree plants and THI-1 utilize the same model of Dresser safety valves. [Ref. 11.3 Further-more, as discussed above, incorrect ring settings can result in inadequate relief capacity and damage to the safety valves. Since the staff claims that the " limiting transient" for TMI-1 is a suberitical rod withdrawal [Ref. 12.]

and such an accident could occur early in the restart process, resolution f this safety issue can not be postponed until after restart,

b. NUREG-0737. Item II.F.2, Attachment 1, Design and Qualification Criteria for Pressurized-Water Reactor Incore Thermocouples.

One of the " lessons learned" requirements was that additional instrumen-tation was needed for detection of inadequate core cooling. One requirement applicable to incore thermocouples that measure core exit temperatures was that "[t]he instrumentation should be environmentally qualified . . .. " [Ref.

13.]

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A GPU submittal to the staff dated May 20, 1983 stated that no extension of the March 31, 1985 deadline in 10 CFR 50.59 for environmental qualification of electrical equipment would be needed. In a submittal dated December 11, 1984, GPU " identified nine (9) items of equipment that were not qualified."

including the incore thermocouple instrumentation. [Ref. 14.] On February 12, 1985, GPU applied for an extension of the deadline for qualifying the thermo-couples. GPU gave as its " reason for scheduler [ sic] extension" the following:

Cualification testing of the existing cables and connectors is planned to be complete in April 1985 based on vendor bids.

The final report is due out in May 1985. Pending the outcome of the testing, replacement may be necessary. GPUN is pro-ceeding in a parallel path with the testing to procure environmentally qualified replacement cables . . ..

[Ref. 15.] , .

GPU offered as a "JIO" (Justification for Interim Operation)2/ the asser-tion that "[t]he safety function (of the incore thermocouple instrumentation 3 can be accomplished by alternate qualified instrumentation with existing j p.ocedural guidance." [Ref. 16.] GPU asserted that the hot leg resistance ,

temperature detectors (RTDs) are the " alternate qualified instr = mentation" to ba used in determining the subcooling margin of the reactor coolant system.

However, GPU also correctly informed the staff, albeit in a footnote, that

"[w]ith reactor coolant pumps running, the subcooling margin is calculated from the hot leg RTDs. If the RCPs are off, the subcooling margin is calcu-lated from the incore thermocouples." [Ref.17.3 2/ Previously, such information has been termed a "JC0" (Justification for Continued Operation). The criteria for an acceptable justification for equipment not snown to be qualified require, inter alia, either a) the equipment would perform its function before failing or b) there is other

. redundant, qualified safety-grade equipment available to perform the same

. safety function.

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. The staff appears to have overlooked GPU's footnote. In a letter to GPU j dated March 29, 1985,E the staff stated that "[w]e have reviewed the JIOs and conclude that they are sufficient to support operation during the requested extension period," and granted .GPU's requested extension of the March 31, 1985 environmental qualification deadline. [Ref. 18.]

However, even assuming, arguendo, that the instrumentation used to determine subcooling margin from the hot led RTDs is environmentally quali-fied, that intrumentation can not substitute for the incore thermocoupies bacause it can not be used unless the reactor coolant pumps are running.

Under design basis accident assumptions, the reactor coolant pumps will not be running. The Commission's regulations (GDC-17) require postulating the loss of offsite electrical power, which is the only source of power for the pumps.

Furthermore, even if offsite power is not lost, the TMI-1 reacto.- operators have been trained and instructed to shut off the reactor pumps in certain accident situations. Thus, UCS concludes that NUREG-0737 Item II.F.2 is not complete because the incore thermocouples are not environmentally qualified. .

UCS also concludes, contrary to the staff, that GPU has not provided a valid justification for operation of 1HI-1 without environmentally-qualified incere thermocouple instrumentation. One or the other must be done before operation of THI-1 can go forward uunder the rules that apply to all plants.

2. GPU Commitments or License Conditions That Are Not Met One of the lessons learned requirements was that provisions to supply emergency electrical power to the pressurizer heaters must be made without degrading the capacity, capability or reliability of the onsite emergency

}/ UCS received a copy of this letter at 4:00 p.m., Friday, April 15, 1985.

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power supply. During the restart proceeding, UCS contended that the provi-sions at INI-1 to protect against the loss of the emergency electrical power supply as a result of an electrical fault (e.g., short circuit) in the non-safety grade pressurizer heaters or their circuits were in' adequate. [Ref. 19.]

This safety issue was resolved by the Boards relying upon, inter alia, GPU's commitments that two conditions would be satisfied at THI-1. First, the Eoards relied upon GPU's testimony that "[t]he undervoltage relays will trip open the [ pressurizer heater circuit] breaker if the voltage on the ES

[ engineered safeguards] bus drops to 430v for 1.5 seconds." [Ref. 20.]

Second, the Boards relied upon GPU's testimony that the pressurizer heaters will not be connected to the emergency power supply unless " adequate diesel generator capacity is available and all systems have stabilized." [Ref. 21.]

GPU apparently has no intention of adhering to these two conditions.

Du.-ing routine safety inspections of TMI-1 conducted on January 3 -

February 3, 1984 NRC inspectors noted that GPU had changed both the voltage and time delay setpoints for the undervoltage trip. The inspectors reported -

that the setpoints were changed from those relied upon by the Appeal Board to avoid tripping of the pressurizer heater circuit breakers "due to motors starting on the class 1E bus." (If motors are being started, then the load on the power supply has noc been stabilized.) Although aware that the Appeal Board decision was based on different setpoints, the inspection report states that "[t]his item is closed." The relevant portion of the inspection report follows:

The setting specified in the Licensee [ sic] Board Hearing ALAB-729 was 430 volts at 1.5 seconds. The setting by the licensee is 420 volts at 1 second.

The ALAB-729 relay setting of 430 volts would protect the class 1E system. However, the purpose of CLI-79-8 would not be achieved due to undesired tripping. This setting is 90% of normal [ voltage] which would be reached due to motors starting

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l on the class 1E bus. The setting of 420 volts is 85% of normal which allows 80% for motor starter coil pickup plus 5%

voltage drop through the control transformer and control wires. The time setting has been reduced from the ALAB-729 requirement of 1.5 seconds to 1 second. The 85% voltage setting at 1 second assures the primary objective of pro-tecting the class 1E system. This item is closed.-[Ref. 22.3 It is not clear what evaluation, if any, was done by the staff to support the conclusion that revised undervoltage trip setpoints achieve "the primary objective" of protecting the onsite emergency power supply. It is cle.ar ,

however, that the revised setpoints are net equivalent to those relied upon by the Appeal Board. For example, an electrical fault that caused the voltage to drop to between 420 and 430 volts would not result in tripping of the circuit breakers no catter how long the fault persisted if the unauthorized setpoints are used. In contrast, the same fault would result in tripping the breaker if the setpoints relied upon by the Appeal Board are used. Thus, to some degree, the revised setpoints degrade the capability and reliability of the emergency power supply. In any event, the staff and CPU can not unilaterally change commitments relied upon by the Boards.

UCS considers the abandonment of the commitment not to connect the pres-surizer heaters to the emergency power supply unless the loads have stabilized to have even more safety significance because a failure of the emergency power syste= at this stage could imperilall of the plant's safety systems. Further-more, if that commitment had not been abandoned, there would apparently have been no need to depart from the specific voltage and time delay setpoints for the undervoltage trip which were relied upon by the Appeal Board.

3. TMI-1 TecP"acal Specifications That Negate Board Decisions A principal purpose of the restart proceeding was to decide, in view of the TMI-2 accident, what new safety equipment was necessary ano sufficient to

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8-protect the health and safety of the public during operation of INI-1 That i

process has been completely frustrated by the approval of license technical specifications that permit operation of THI-1 with equipment inoperable, even l though the Eoards found that equipment to be necesstry to protect the public.

Examples of instances where this has occurred are as follows:

a. Subcooling Margin Instrumentation. A lessons learned requi.ement was that primary coolant saturation meters should be installed to provide online indication of coolant saturation condition. [Ref. 23.] GPU testified that this requirement was met by the installation two "subcooling margin monitors,"

one for each loop. However, the technical specification permit unrestricted operation of IMI-1 with one subcooling meter inoperable all the time and permit operation for an additional week after failure of the second. [Ref. 24.3 The purported basis for tnis technical specification is that "[a)1 ternate indications are available for Saturation Margin Monitors using hand calcula-tions." [Ref. 25.3 However, the use of " hand calculations," wnich were avail-eble during the INI-2 accident, obviously does not meet the lessons learned requirement to install instrumentation "to provide on-line indication of coolant saturationcondition." [Ref. 26.) Furthermore, the only temperature indications available for such hand calculations, other than the hot leg resistance temperature detectors which provide the temperature input to the subcooling meters, are the incore thermocouples. As discussed above, the staff has p,roposed allowing INI-1 to operate without environmentally qualified incore thermocouple instrumentation on the grounds that the incore thermocouples are backed up by the subcooling meters. In any event, the lessons learned requirements specified that both the incore thermocouples and saturation meters must be provided and GPU tesimony during the restart i

proceeding was'that both would be.

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b. Direct Indication of PORV and Safety Valve Position. A lessons learned requirement was that reactor system relief and safety valves shall be provided with reliable positive indication of their position in the control room. [Ref. 27.] GPU testified that this requirment had been met by the installation of redundant position indicators for the PORY and by a single position indicator for each safety valve. Specifically, discharge flow from the two pressurizer code safety valves and the PORV is measured by differen-tial pressure transmitters connected across elbow taps downstream of e'ach valve. In addition, an acoustic monitor is provided to detect flow in the PORV discharge line.

4 However, the technical specifications permit operation of TMI-1 with

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either one of the PORY instruments continuously inoperable and permit con-tinued operation for a week following the failure of the other PORV instrument

- and both safety position indications. [Ref. 28.3 The basis purported to justify this relaxation of the lesson learned requirements is that alternate ~

indication of the position of the safety valves and the PORY is available from l

the thercocouples in the valves' discha ge pipes. It is precisely these downstream thermocouples that proved inadequate during the TMI-2 accident and l

I led to the requirement for reliable position indication. ,_

c. Emergency Feedwater System Flow Indication. A lessons learned requirement was that "[s]afety-grade indication of auxiliary feedwater flow to each steam gererator shall be provided in the control room." [Ref. 29.3 GPU testified that this reouirement was met by the installation of redundant EFW i

l flow instruments for each steam generator.

l l However, the technical specifications permit operation of TMI-1 with one l

l flow instrument for each steam generator inoperable all the time and permit cantinued operation for a week -following failure cf the other EFW flow instro-

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a msnts. [Ref. 30.) Allowing operation with one flow instrument inoperable does not meet the requirement for si.fety-grade instrumentation because the single failure criterion is not met with only one flow indicator per steam generator.

The availability of local EFW flow indication does not satisfy the requ.'rement that the indication be in the control room,

d. Reactor Coolant System High Point Vents. A lessons learned require-ment was that high points vents be installed inorder to vent non-condensible gases from the reactor coolant system which could otherwise block natural circulation cooling of the core. The lessons learned requirements applicable to the high point vents specified that although each vent path need not meet the single failure criterion, a " degree of redundancy should be provided by powering different vents from different emergency buses." [Ref. 31.]

The evidentiary record indicates that IMI-1 complied with this require-ment by installing four vents -- one on each loop, one on the pressurizer and one on the reactor vessel. The design is such that the vents on the vessel, pressurizer, and loop B are powered f.om one electric power supply and the vent on loop A is powered from a redundant power supply.

However, the technical specifications pe mit operation of TMI-1 with one loop vent inoperable all the time and permit continued operation for 30 days I

following failure of another vent. [Ref. 32.] Thus, if 7NI-1 is operating i with the loop A vent inoperable, a single failure of the power supply for the

[

pressu5izer, vessel and ' loop B vents would render all high point vents inoper-able. The technical specifications therefore violate the lesson learned rtquirement that the vents collectively should meet the single failure criter-ion. Furthermore, since the vents on the pressurizer and the vessel can not r; move gas from the high point - of the loops, the same situation (i.e., the inability to restore natural circulation) would prevail if IMI-1 were in oper-

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ation with the loop B vent inoperable. That is, an accident accompanied by a single failure of the power supply for the loop A vent would leave TMI-1 with no operable loop vent. The basis for this technical specification provides no justication either .for . allowing operation with one loop vent continuously inoperable, or for 30-days continued operation with both loop vents inoperable.

UCS concludes that the IMI-1 Technical specifications must be amended to require, as " limiting conditions for operation," the operability of all equip-cant relied upon by the Boards as a basis for the decision that D1I-1 can be operated without endangering the health and safety of the public. In addition, we believe that another requirement should be added to the technical specifications.

The technical specifications permit continued operation for some period of time following failure of safety equipment that is subject to a limiting condition for operation. If repairs can not be made with the specified time, the technical specification require that the plant be shut down. UCS believes .

that GPU should be required to assess, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after detecting the failure, whether the equipment can be restored to an operable status within the time continued operation is permitted without that equipment. When - it -i s concluded that repairs can not be accomplished within the allowable time p2riod, the plant should be immediately shut down. There is no justification for continued operation, even for a limited time, if it is known that the equipment can not be repaired prior to the time the plant vill be required to shut down.

4. References. See next page. ~

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References

1. NUREG-0737. Item II.D.1. See also, NUREG-0578, Section 2.1.2.

2. Appeal Tr. pp. 151-152 (Lanese), March 7, 1983. -

3. J. H. Taylor, Manager, Licensing Services, B&W, letter to R. C. DeYoung, Director, IE, NRC, October 15, 1982, p. 1.

4 See LBP-81-59, 14 NRC 1211, 1375 (1981).

5. 14 NRC 1378-79. .

6. Jones and Lanese, fol. Appeal Tr. 111 at 3-4, March 7, 1983, emphasis added.

7. Id., at 4

8. Appeal Tr. 423 (Correa), March 16, 1983, emphasis added.

9, NRC Inspection Report No. 50-239/85-03, March 18,1985, p. 3.

10. Frank Miraglia and Gus Lainas, Division of Licensing, NRC, memorandum to James Knight, Division of Engineering, NRC, " Review of Licensee / Vendor Letter on Dresser Safety Valves," November 3,1982.

11. Thocas Ippolito, Division of Licensing, NRC, memorandum to Dennis Crutchfield, Division of Licensing, NRC, " Review of Licensee / Vendor Letters on Dresser Safety Valves," December 30, 1982, enclosure, Table 1. .

12. Id.

13. NUREG-0737, Item II.F.2, Attachment 1, p. 3-118.

14 Harold Denton, Director, NRR, NRC, letter to Henry Hukill, Vice President and Director - TMI-1,.GPUN, March 29, 1985, p. 1.

15. GPUN letter no. 5211-85-2018 to Harold Denton, NRC, February 12, 1985,

p. 1, emphasis added.

16. Id., enclosure, p. 6..

17. Id., e6 closure, p. 3, n. 1.

, 18. Harold Denton, Director, NRR, NRC, letter to Henry Hukill, Vice President

! and Director - TMI-1, GPUN, March 29, 1985, p. 2.

19. See , e.g. , ALAB-729, 17 NRC 814, 855-859 (1983).

20. 17 NRC 858, footnote omitted.

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21. 17 NRC B60, footnote omitted, emphasis added.

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22. NRC Inspection Report No. 50-289/84-01, March 17, 1984 (NRC transmittal letter datec April 9, 1984), p. 5.

23. NUREG-0578, pp. . A-11, A-12. .

24 Technical Specification 3.5.5 (Amendment No. 100) and Table 3.5-2 (Amendment No. 78). ,

25. Technical Specification 3.5.5, p. 3-40b ( Amendment No.100).

26. NUREG-0578,-p. A-12.

27. NUREG-0578, pp. A-9, A-10.

28. Technical Specification 3.5.5 (Amendment No. 100) and Table 3.5-2 -

(Amendment No. 78)..

29. NUREG-0578, p. A-32.

30. Technical Specification 3.5.5 ( Amendment No.100) and Table 3.5-2 (Amendment No. 78).

31 NUREG-0737, Item II.B.1,~ Reactor Coolant System Vents.

32. Technical Specification 3.1.13 (Amendment No. 97).

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UNION OF CONCERNED SCIENTISTS im c...... ..... s.w. . s. noi . msi.,1.. oc 2oose . <202> 2,6.seoo 00( ME7t t, USHRC May 3, 1985 T3 MAY -6 All :52 Nun::io J. Palladino, Chairman Thac:as M. Roberts, Cocrnissioner' 0FFICE]&c:gg[g[h.

00ChE7 sG's c James K. Asselstine, Cocznissioner dRANCH Frederick M. Bernthal, Coccissioner Lando W. Zech, Commissioner U. S. Nuclear Regulatory Ccx:rnission ~

Washington, D. C. 20555 Gentlemen:

On April 17, 1985, the NRC staff briefe<f the Ccx:rnission on the status of Three Mile Island Unit 1 (NI-1) in preparation for a restart vote. UCS's ccments the following day on the staff's presentation were limited to the subject of the steam generators. We are writing now to address other subjects because the staff's presentation was not complete.

Specifically, the staff's presentation failed to inform you of the following:

1. Scx::e " lessons learned" items (NUREG-0737 items) previously certified by the staff to be cceplete are, due to new information, no longer complete;

2. Sece conditions of the WI-1 plant do not conform with the conditions

described in sworn testimony during the restart proceeding which were relied upon by the Licensing Board and Appeal Board in deciding contested issues in favor of restart; and

3. Many of the new TMI-1 technical specifications proposed by GPU and approved by the staff effectively reverse Licensing Board and Appeal Board findings that certain equipment is necessary to protect public health and safety by pemitting operation of THI-1 with that equipment inoperable.

Specific e::amples of each of these safety deficiencies are desedbed in detail in the attachment to this letter. A brief sunr::ary of those examples follows.

1.a.' A lessons learned requirement was that the pressurizer safety valves be tested to demonstrate acceptable perfomance under accident conditions.

After tests were brought to the Appeal Board's attention showing that inc.ve-rect valve adjustments caused valve damage and inadequate relief capacity, GPU ccmitted to specific valve settings that had been shown to produce acceptable valve performance. 'Ihe issue was largely resolved on this basis. However, GPU later changed the settings, as discoverd by an NRC inspector. Documenta-tion that purportedly supports the new settings is dated in early 1984 but was not scheduled to be submitted by GPU until March 31, 1985. UCS has not received that' infomation or the staff's evaluation of it.

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l' detect1.b. The 1NI-2 inadequate accident core coolingdemonstrated was required. that additional instrumentation to The incore thermocouples are part of that instrumentation and a lessons learned requirement was that the incere thennocouple instrumentation must be environmentally qualified. GPU informed the staff that environmental qualification has not been established testing is underway, efforts to procure replacement equipment are being pur, sued concurrently, and if replacement is necessary, procurement and installa-i tion will take 40 weeks. On March 29, 1985, the staff approved GPU's request for extension of the March 31, 1985 environmental qualification deadline on the basis that the hot leg temperature detectors (which provide an input signalthermocouples.

incere to the subcooling margin meters) are an adequate substitute for the However, it was apparently overlooked that the hot leg temperature instruments can not be used unless the reactor coolant pumps are running, which they will not be in certain accident conditions. In addi' tion,-

t the 1MI-1 technical specifications permit plant operation with the subcooling margin instrumentation are an adequate substitute. inoperable on the basis that the incore thermocouples

2. Another emergency lesson electrical learned requirement was that provisions to supply degrading the capacity, power to the pr_esurizer heaters must be made without power supply. capability or reliablility of the onsite emergency The boards found the 1MI-1 provisions to be acceptable in large part on the basis that two conditions would be satisfied: a) the heaters would not and beb) connected until the load on the emergency power supply was stabilized; that the undervoltage trip used to disconnect the heaters to protect the power supply from a failure in the heater circuits would be set at a specified voltage and time delay. GPU has since changed the setpoints for both the voltage and time delay on the basis that starting of motors on the emeegency power supply would result in tripping the heaters. This indicates that the power heaters will be connected without first stabilizing the load on the supply. ,

The staff other than to classify it asisaaware " closed of item."

this situation but has taken no action l

3. A principal purpose of the restart proceeding was to decide, in view of to the 1MI-2 accident, what new safety equipment was necessary and sufficient protect That processthe health and safety of the public during operation of TMI-1.

has been completely frustrated by the approval of technical

!. specifications that permit operation of 1MI-1 with equipment inoperable which I

the boards found was necessary to protect the public. For example, the following new safety equipment falls in this category: subecoling margin inr.trumentation, position indication for the pressurizer PORY and safety

valves, system highemergency feedwater system flow instruments, and reactor coolant point vents.

UCS believes that the Comission should take the following actions:

1. Direct the staff to identify any additional " lessons learned" require-ments but which or other are noitems longerthat the staff has certified to the Comssion as complete complete.

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2. Direct the staff to identify any additional instances where the present status of TMI-1 does not conform with the GPU commitment and/or board-imposed license condition that was relied upon in the adjudicatory decisions favorable to restart.

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method Forand items 1. and 2. above, the staff should also be directed to propose a schedule for resolving each safety issue identified, giving due consideration to the rights of the other parties. i

3. Direct the staff to review the TMI-1 technical specifications and a end the= as necessary to require, as " limiting conditions for operation,"

the operability of all equi;xnent relied upon by the boards as a basis for decision of the public. that TMI-1 can be operated without endangering the health and safety In addition, the staff should be directed to serve the parties with a complete set of the amended technical specifications and solicit the parties' conrnents on whether the amended technical specifications contain all such required limiting conditions.for operation.

In sum, UCS believes that there are more items requiring resolution prior to a Consnission vote on restart than were identified by the staff in its April 17, 1985 presentation to the Ccanission. We anticipate the charge that these questions are being raised at the last minute. On the contrary, these are not new questions.

They are issues which were purportedly settled during the restart proceedings but which have becczne unraveled by either inadvertent or intentional lack of attention to the conditions of restart imposed by or i=plicit in the ASLB and Appeal Board decisions.

Sincerely, 1

Robert D. Pollard Nuclear Safety Engineer s .

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Ellyn . Weiss General Counsel enel: Additional Safety Issues Requiring -

Resolution Prior To TMI-1 Restart ec: TMI-1 Service List 1

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ADDITICNAL SAFETY ISSUES REQUIRING RESOLUTION PRIOR TO TMI-1 RESTART Union of Concerned Scientists DOCKET _E0 U5n.%

May 3, 1985

'85 tM-6 All :52

1. " Lessons Learned" Recuirements That Are Incomplete

a. hUREC-0737, Item II.D.1, Performance Testing of soilihs&d,CEa G7 H EEC'ET:.'

(E'ehF 2 -

Pressurized-Water Reactor Relief and Safety Valves (NUREG-0578, Section 2.1.2)

One of the lessons learned requirements was that reactor coolant system safety valves had to be tested under the full range of expected operating con-ditions for design basis transients and accidents, which would include solid-wster and two-phase flow conditions. [Ref. 1.3 As discussed below, data from testing Dresser Industry safety valves, including the specific model used as pressurizer safety valves at IMI-1, showed that certain plant modifications and adjustments 1/ to the safety valves were necessary to prevent valve chttter that could reduce relief capacity and damage the valve internals. GPU commit-

'.s d to these changes, which were relied upon by the Appeal Board to resolve .

the issue.

Recent information indicates that CPU will not utilize the valve settings relied on by the Eoard.

1/ Tnese adjustments are called ring settings. Although there is some inter-l action between the three separate ring settings, the principal effect of l

each ring setting is es follows: the lower ring setting affects the rate l at which the valve opens and adjusts " accumulation," the pressure above

' the preset opening pressure at which the valve is fully open; the middle ring adjusts " blowdown," the pressure below the opening pressure at which the valve is fully reclosed; the upper ring setting is used to compensate for the effects of backpressure downstream of the valve. Accumulation and tiowdown are generally expressed as a percentage of the pressure setpoint.

For example, a safety valve set to open at 2500 psi with 3% accumulation and 101atblowdown 2250 psi.should open at 2500 psi, be fully open at 2575 psi and l .

reclose (Ref. 2.]

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B&W provided the 'following evaluation of the results from the testing performed vi the safety valves:

Recent review of the EPRI test data indicate that for certain combinations of lower and middle ring settings. the safety valves could chatter (rapid cycle valve opening and closing).

Two consequences of valve chattering have been noted during valve testing: a reduction in safety valve capability to relieve pressure while chattering and subsequent damage to valve internals. During the EPRI tests, this chattering phenomenon was noter,' only with valves with long inlet piping configurations. However, present understanding of the effect ,

, of ring settings on valve operation indicates that even with the valves moun::ed directly on the pressurizer nozzles, valve chatter could be postulated for a combination of ring settings significantly different from those used during the EPRI tests.

[Ref. 3.]

During the restart hearings conducted by the Licensing Board, UCS assert-ed, inter ' alia, that appropriate qualification testing of the safety valves had not been performed to verif y th- capability of these valves to function during normal, transient, and accident conditions. [Ref. 4.3 In resolving this question, the Licensing Board found that a valve testing program being conducted by the Electric Power Research Institute (EPRI) was adequate to reveal any design deficiencies in the safety valves used at TMI-1. [Ref. 5.]

This subject was revisited during the reopened hearings held by the Appeal Board in March 1983. ??U testifiec that as a result of the EPRI tests

"[t]he safety valves were shown to be acceptable under all modes of operation i

after certain modifications were made." [Ref. t..] The modifications which GPU testified were necessary in order to assure acceptable safety valve perform-ance were to change "the safety valve inlet piping from a long inlet to a short inlet arrangement" [Ref. 7.], and to readjust the safety valve ring set-tings to "the final EPRI ring settings which we used in the last five tests."

, [Ref. 8.] GPU's commitment'to the latter modification has been abandoned.

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, Duri.ng an NRC inspection of INI-1 conducted on January 21-24, 1985 (but not reported until March 18, 1985), the inspector noted that GPU had changed

the safety valve ring settings. CPU furnishe'd the inspector with documenta-I tion from B&W dated in February and May of 1984 that purportedly supports the ,

revised ring settings. However, the inspector concluded that since the staff's safety evaluation report had been based on the earlier ring settings, there was a need to inform the Office of Nuclear Reactor Regulation (NRR) of the details. GPU representatives agreed "and committed to furnish the r-ing setting information to NRR by 3/31/85." [Ref. 9.] To date, UCS has not received that info mation or any staff evaluation of it.

j UCS believes that this safety issue must be resolved before restart. The i

ssfety significance of the ring settings for the pressurize safety valves is illustrated by the fact that after reviewing their safety valve ring settings

, against the EPRI test results, Duke Power Company declared the safety valves inoperable and shut down Oconee Unit 2. [Ref. 10.] The Oconee plants and IMI-1 utilize the same model of Dresser safety valves. [Ref. 11.3 Further-more, as discussed above, incorrect ring settings can result in inadequate relief capacity and damage to the safety valves. Since the staff claims that the " limiting transient" for INI-1 is a suberitical rod withdrawal [Ref. 12.]

and such an accident could occur early in the restart process, resolution of this safety issue can not be postponed until after restart.

b. NUREG-0737. Item II.F.2, Attachment 1, Design and Qualification Criteria for Pressurized-Water Reactor Incore Thermocouples.

One of the " lessons learned" requirements was that additional instrumen-tation was needed -for detection of inadequate core cooling. One requirement applicable to incore thermocouples that measure core exit temperatures was that "[t]he instrumentation should be environmentally qualified . . .. " [Ref.

13.]

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A GPU submittal to the staff dated May 20, 1983 stated that no extension of the Maren 31, 1985 deadline in 10 CFR 50.59 for environmental qualification of electrical equipment would be needed. In a submittal dated December 11, 1984, GPU " identified nine (9) items of equipment that were not qualified,"

including the ir. core thermocouple instrumentation. [Ref. 14.3 On February 12, 1985, GPU applied for an extension of the deadline for qualifying the thermo-couples. GPU gave as its " reason for scheduler [ sic] extension" the following:

Cualification testing of the existing cables and connectors

' is planned to be complete in April 1985 based on vendor bids.

The final report is due out in May 1985. Pending the outcome of the testing, replacement may be necessary. GPUN is pro-4 i ceeding in a parallel path with the testing to procure environmentally qualified replacement cables . . . .

[Ref. 15.3 -

I GPU offered as a "JIO" (Justification for Interim Operation)- the asser-i tion that "[t]he safety function (of the incore thermocouple ' instrumentation]

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i can be accomplished -by alternate qualified instrumentation with ex'isting I

procedural guidance." [Ref. 16.3 GPU asserted that the het{ 1eg resistance ,

tamperature detectors (RTDs) are the " alternate qualified instr.unentation" to ba used in determining the subcooling ma. gin of the reactor coolant system.

However, GPU also correctly informed the staff, albeit in a footnote, that

"[w]ith reactor coolant pumps running, the subcooling margin is calculated from the hot leg RTDs. If the RCPs are off, the subcooling margin is calcu-Icted from the incore thermocouples." [Ref.17.3 y Previously, such information has been termed a "JC0" (Justification for Continued Operation). The criteria for an acceptable justification for equipment not shown to be ~ qualified ' require, inter alia, either a) the equipment would perform its function before failing or b) there is other redundant, qualified safety-grade equipment available to perform the same

- safety function.

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The staff appears to have overlooked CPU's footnote. In a letter to GPU dated March 29, 1985,1 the staff stated that "[w]e have reviewed the JIOs and conc 1;de that they are sufficient to support operation during the requested extension period," and granted GPU's requested extension of the March 31, 1985 environmental qualification deadline. [Ref. 18.)

However, even assuming, arguendo, that the instrumentation used to determine subcooling margin from the hot led RIDS is environmentally quali-fled, that intrumentation can not substitute for the incere thereccou' des btcause it can not be used unless the reactor coolant pumps are running.

Under design basis accident assumptions, the reactor coolant pumps will not be running. The Commission's regulations (GDC-17) require postulating the loss

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of offsite electrical power, which is the oniv source of power for the pumps.

Furthermore, even if offsite power is not lost, the TNI-1 reacto.- operators have been trained and instructed to shut off the reactor pumps in certain cecident situations. Thus, UCS concludes that HUREC-0737 Item II.F.2 is not complete because the incore thermocouples are not environmentally qualified. .

UCS also concludes, contrary to the staff, that GPU has not provided a valid justification for operation of THI-1 without environmentally-qualified incore thermocouple instrumentation. One or the other must be done before ope. ation of INI-1 can go forward uunder the rules that apply to all plants.

2. GPU Commitments or License Conditions That Are Not Met One of the lessons learned requirements was that provisions to supply smargency electrical power to the pressurizer heaters must be made without dagrading the capacity, capability or reliability of the onsite emergency i

l 1/ UCS received a copy of this letter at 4:00 p.m., Friday, April 15, 1985.

l

power supply. During the restart proceeding, UCS contended that the provi-sions at WI-1 to protect against the loss of the emergency electrical power supply as a result of an electrical fault (e.g., short circuit) in the non-l safety grade pressurizer heaters or their circuits were inadequate. [ Ref. 19. ]

l This safety issue was resolved by the Boards relying upon, inter alia,

! GPU's commitments that two conditions would be satisfied at TMI-1. First, the Boards relied upon CPU's testimony that "[t]he undervoltaEe relays will trip open the [ pressurizer heater circuit] breaker if the voltage on the ES

[ engineered safeguards] bus drops to 430v for 1.5 seconds." [Ref. 20.]

Second, the Boards relied upon GPU's testimony that the pressurizer heaters will not be connected to the emergency power supply unless " adequate diesel generator capacity is available and all systems have stabilized." [Ref. 21.]

GPU apparently has no intention of adhering to these two conditions.

During, routine safety inspections of EI-1 conducted on January 3 -

February 3, 1984, NRC inspecto. s noted that GPU had changed both the voltage and time delay setpoints for the undervoltage trip. The inspectors reported -

that the setpoints were changed from those relied upon by the Appeal Board to avoid tripping of the pressurizer heater circuit breakers "due to motors starting on the class 1E bus." (If motors are being started, then the load"on the power supply has not been stabilized.) Although aware that the Appeal Board decision was based on different setpoints, the inspection report states tnat "[t]his item is closed." The relevant portion of the inspection report follows:

The setting specified in the Licensee (sic] Board Hearing ALAB-729 was 430 volts at 1.5 seconds. The setting by the licensee is 420 volts at 1 second.

l The ALAB-729 relay setting of 430 volts would protect the class 1E system. However, the purpose of CLI-79-8 would not i be achieved due to undesired tripping. This setting is 90% of I

normal [ voltage] which would be reached due to motors starting

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l on the class 1E bus. The setting of 420 volts is 85% of normal which allows 80% for motor starter coil pickup plus 5%

voltage drop through the control transformer and control wires. The time setting has been reduced from the ALAB-729 requirement of 1.5 seconds to 1 second. The 85% voltage setting at 1 second assures the primary objective of pro-tecting the . class 1E system. This item is closed.-[Ref. 22.)

It is not clear what evaluation, if any, was done by the staff to support the conclusion that revised undervoltage trip setpoints achieve "the primary objective" of protecting the onsite emergency power supply. It is cigar, however, that the revised setpoints are not equivalent to those relied upon by the Appeal Boarc. For example, an electrical fault that caused the voltage to drop to between 420 and 430 volts would not result in tripping of the circuit breakers no matter how long the fault persisted if the unauthorized setpoints are used. In contrast, the same fault would result in tripping the breaker if the setpoints relied upon by the Appeal Board are used. Thus, to some degree, the revised setpoints degrade the capability and reliability of the emergency power supply. In any event, the staff and GPU can not unilaterally change commitments relied upon by the Boards.

UCS considers the abandonment of the commitment not to connect the pres-surizer heaters to the emergency power supply unless the loads have stabilized to have even more safety significance because a failure of the emergency poYer system at this stage could imperilall of the plant's safety systems. Further-more, if that commitment had not been abandoned, there would apparsntly have baen no need to depart from the specific voltage and time delay setpoints for the undervoltage trip which were relied upon by the Appeal Board,

3. TM1-1 Technical Specifications That Negate Board Decisions A principal purpose of the restart proceeding was to decide, in view of the THI-2 accident, what new safety equipment was necessary and sufficient to

protect the health and safety of the public during operation of THI-1. That process has been completely frustrated by the approval of license technical specifications that permit operation of IMI-1 with equipment inoperable, even though the Eoards found that equipment to be necessary to protect the public.

Eramples of instances where this has occurred are as follows:

a. Subcooling Margin Instrumentation. A lessons learned requirement was that primary coolant saturation meters should be installed to provide online indication of coolant saturation condition. [Ref. 23.3 CPU testified that this requirement was met by the installation two "subcooling margin monitors,"

one for each loop. However, the technical specification permit unrestricted operation of IMI-1 with one 'subcooling meter inoperable all the time and permit operation for an additional week after failure of the second. [Ref. 24.3

. The purported basis for this technical specification is that "[allternate indications are available for Saturation Margin Monitors using hand calcula-tions." [Ref. 25.] However, the use of " hand calculations," which were avail-able durinE the IMI-2 accident, obviously does not meet the lessons learned requirement to install instrumentation "to provide on-line indication of coolant saturationcondition." [Ref. 26.3 Furthermore, the only temperature indications available for such hand calculations, other than the hot leg resistance temperature detectors which provide the temperature input to the subcooling meters, are the incore thermocouples. As discussed above, the staff has p,roposed allowing TMI-1 to operate without environmentally qualified incore thermocouple instrumentation on the grounds that the incore j thermocouples are backed up by the subcooling meters. In any event, the lessons learned requirements specified that both the incore thermocouples and saturation meters must be provided and GPU tesimon/ during the restart proceeding was that both would be.

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b. Direct Indication of PORV and Safety Valve Position. A lessons Isarned requirement was that reactor system relief and safety valves shall be provided with reliable positive indication of their position in the control room. [Ref. 27.3 GPU testified that this requirment had been met by the installation of redundant position indicators for the PORV and by a single position indicator for each safety valve. Specifically, discharge flow f.-om the two pressurizer code safety valves and the PORV is measured by differen-

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tial pressure transmitters connected across elbow taps downstream of each valve. In addition, an acoustic monitor is provided to detect flow in the PORV discharge line.

However, the technical specifications permit operation of INI-1 with either one of the PORV inst ~ruments continucusly inoperable and permit con-tinued operation for a week following the failure of the other PORV instrument and both safety position indications. [Ref. 28.) The basis purported to I

justify this relaxation of the lesson learned requirements is that alternate indication of the position of the safety valves and the PORV is available from the thermocouples in the valves' discha ge pipes. It is precisely these d:wnstream thermocouples that proved inadequate during the TMI-2 accident and led to the requirement for reliable position indication. _

c. Emergency Feedwater System Flow Indication. A lessons learned requirement was that "[s]afety-grade indication of auxiliary feedwater flow to ecch steam generator shall be provided in the control room." [Ref. 29.3 GPU testified that this requirement was met by the installation of redundant EFW flow instruments for each steam generator.

However, the technical specifications permit operation of TMI-1 with one flow instrument for each steam . generator inoperable ' all the time and permit I

continued operation for a week following failure of the other EFV flow instru-

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mants. [Eef. 30.] Allowing operation with one flow instrument inoperable does not meet the requirement for safety-grade instrumentation because the single failure criterion is not met with only one flow indicator per steam generator.

The availability of local EFW flow indication does not satisfy the requirement that the indication be in the control room,

d. Reactor Coolant System High Point Vents. A lessons learned require-mont was that high points vents be installed inorder to vent non-condensible gases from the reactor coolant system which could otherwise block natural circulation cooling of the core. The lessons learned requirements applicable to the high point vents specified that although each vent path need not meet the single failure criterion, a " degree of redundancy should be provided by powering different vents from different emergency buses." [Ref. 31.3 The evidentiary record indicates that IMI-1 complied with this require-nent by installing four vents - one on each loop, one on the pressurizer and one on the reactor vessel. The design is such that the vents on the vessel, pressurizer, and loop B are powered f.om one electric power supply and the vent on loop A is powered from a redundant power supply.

However, the technical specifications pe mit operation of IMI-1 with one loop vent inoperable all the time and permit continued operation for 30 da'ys following failure of another vent. [Ref. 32.) Thus, if 2NI-1 is operating !

l with the loop A vent inoperable, a single failure of the power supply for the i

. \

pressurizer, vessel and ' loop B vents would render all high point vents inoper-cble. The technical specifications therefore violate the lesson learned requirement that the vents collectively should meet the single failure criter-ion. Furthe more, since the vents on the pressurizer and the vessel can not rsmove gas from the high point of the loops, the same situation (i.e., the inability to restore natural circulation) would prevail if 7NI-1 were in oper-

. ~

stion with the loop B vent inoperable. That is, an accident accocipanied by a single failure of the power supply for the loop A vent would leave INI-1 with no operable loop vent. The basis for this technical specification provides no justication either for . allowing operation with one loop vent continuously inoperable, or for 30-days continued operation with both loop vents inoperable, f

UCS concludes that the TMI-1 Technical specifications must be amended to i

require, as " limiting conditions for operation," the operability of all equip-mant relied upon by the Boards as a basis for the decision that IMI-1 can be operated without endangering the health and safety of the public. In addition, we believe that another requirement should be added to the technical specifications.

The technical specifications permit continued operation for so:ne period of time following failure of safety equipment that is subject to a limiting condition for operation. If repairs can not be made with the specified time, J

the technical specification require that the plant be shut down. UCS believes .

that GPU should be required to assess, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after detecting the failure, whether the equipment can be restored to an operable status within the time continued operation is permitted without that equipment. When it-is 4

concluded that repairs can not be accomplished within the allowable time i

, pariod, the plant should be immediately shut down. There is no justification for continued operation, even for a limited time, if it is known that the equipment can not be repaired prior to the time the plant will be required to shut down.

)

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4. References. See next page.

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_ 12 References

1. NUREG-0737, Item II.D.1. See also, NUREG-0578 Section 2.1.2.

2. Appeal Tr. pp. 151-152 (Lanese), March 7, 1983. -

3 J. H. Taylor, Manager, Licensing- Services, B&W, letter to R. C. DeYoung, Director, IE, NRC, October 15, 1982, p. 1.

4 See LEP-81-59, 14 NRC 1211, 1375 (1981).

5. 14 NRC 1378-79. .

6. Jones and Lanese, fol. Appeal Tr. 111 at 3-4, March 7, 1983, emphasis added.

7. Id., at 4

! 8. Appeal Tr. 423 (Correa), March 16, 1983, emphasis added.

9. NRC Inspection Report No. 50-289/85-03, March 18,1985, p. 3.

10. Frank Miraglia and Gus Lainas, Division of Licensing, NRC, memorandum to James Knight, Division of Engineering, NRC, " Review of Licensee / Vendor Letter on Dresser Safety Valves," November 3,1982.

11. Thomas Ippolito, Division of Licensing, NRC, memorandum to Dennis Crutchfield, Division of Licensing, NRC, " Review of Licensee / Vendor Letters on Dresser. Safety Valves," December 30, 1982,. enclosure, Table 1. .

12. Id.

J

13. NUREG-0737, Item II.F.2, Attachment 1, p. 3-118.

14 Harold Denton, Director, NRR, NRC, letter to Henry Hukill, Vice President and Director - TMI-1,.GPUN, March 29, 1985, p. 1.

15. GPUN letter no. 5211-85-2018 to Harold Denton, NRC, February 12, 1985,

p. 1, emphasis added.

16. Id., enclosure, p. 6. .

17. Id., enclosure, p. 3, n. 1.

18. Harold Denton, Director, . NRR, NRC, letter to Henry Hukill, Vice President and Director - TMI-1, GPUN, March 29, 1985, p. 2.

4 19. See, e.g. , ALAB-729, 17 NRC S14, 855-859 (1983).

( 20. 17 NRC 858, footnote omitted.

t .

21 17 NRC 860, footnote omitted, emphasis added.

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. . 22.' NRC* Inspection Report No. 50-289/84-01, March 17, 1984 (NRC transmittal letter dated April 9, 1984), p. 5.

23. NUREG-0578, pp. A-11, A-12. .

24 Technical specification 3.5.5 (Amendment No.100) and Table 3.5-2 (Amendment No. 78). -

25. Technical specification 3.5.5, p. 3-40b ( Amendment No.100).

26. NUREG-0578, p. A-12.

27. NUREG-0578, pp. A-9, A-10,

28. Technical specification 3.5.5 (Amendment No.100) and Table 3.5-2 -

(Amendment No. 78).

29. NUREG-0578, p. A-32.

30. Technical specification 3.5.5 (Amendment No. 100) and Table 3.5-2 l (Amendment No. 78).

31 NUREG-0737 Item II.B.1,' Reactor coolant system vents.

32. Technical specification 3.1.13 (Amendment No. 97).

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UNION OF CONCERNED' SCIENTISTS 13u c... . ...a.w. . s. noi . w..si. .. oc 2oose . <202> 296.seoo 00( MEit t.

USNRC May 3, 1985 '85 HAY-6 An :52 Nunzio J. hlladino, Chaiman e Thomas M. Roberts, Comissioner' 0FF!c[js.e:ggM$[

James K. Asselstine, Ccmissioner 00CKEi d c ,

dRANCH Frederick M. Bernthal, Coccissioner Lando W. Zech, Cc m issioner U. S. Nuclear Regulatory Ccmnission

Washington, D. C. 20555 Gentlemen:

On April 17, 1985, the NRC staff briefed the Ccmission on the status of Three Mile Island Unit 1 (TMI-1) in preparation for a restart vote. UCS's ecments the following day on the staff's presentation wre limited to the subject of the steam generators. We are writing now to address other subjects because the staff's presentation was not complete.

Specifically, the staff's presentation failed to inform you of the following:

1. Scce " lessons learned" items (NUREG-0737 items) previously certified by the staff to bs complete are, due to new information, nc longer complete;

2. Socie conditions of the TMI-1 plant do not conform with the conditions

described in sworn testimony during the restart proceeding which were relied upon by the Licensing Board and Appeal Board in deciding contested issues in favor of restart; and

3. Many of the new 1MI-1 technical specifications proposed by GPU and approved by the staff effectively reverse 1.icensing Board and Appeal Board findings that certain equipment is necessary to protect public health and safety by permitting operation of IMI-1 with that equipment inoperable.

Speci.~ic examples of each of. these safety deficiencies are described in detail in the attachment to this letter. A brief sumary of those examples follows. .

1.a.' A lessons learned requirement was that the pressurizer safety valves be tested to demon t ate acceptable perfomance under accident conditions.

After tests were brought to the Appeal Board's attention showing that inc.or-rect valve adjustments caused valve damage and inadequate relief capacity, GPU comitted to specific valve settings that had been shown to produce acceptable valve performance. The issue was largely resolved on this basis. However, GPU later changed the settings, as discoverd by an NRC inspector. Documenta-tion that purportedly supports the new settings is dated in early 1984 but was not scheduled to be submitted by GPU until March 31, 1985. UCs has not received that' infomation or the staff's evaluation of it.

1 1.b. The TMI-2 accident demonstrated that additional instrumentation to detect inadequate core cooling was required. The incore thermocouples are part of that instrumentation and a lessons learned requirement was that the incere thermocouple instrumentation must be environmentally qualified. GPU informed the staff that environmental qualification has not been established, testing is underway, efforts to procure replacement equipment are being pur-sued concurrently, and if replacement is necessary, procurement and installa- i i

tion will take 40 weeks. On March 29, 1985, the staff approved GPU's request l for extension of the March 31, 1985 environmental qualification deadline on the basis that the hot leg temperature detectors (which provide an input signal to the subcooling margin meters) are an adequate substitute for the incere thermocouples. However, it was apparently overlooked that the hot leg temperature instruments can not be used unless the reactor coolant pumps are running, which they will not be in certain accident conditions. In addi~ tion,-

the 7MI-1 technical specifications permit plant operation with the subcooling margin instrumentation inoperable on the basis that the incore thermocouples are an adequate substitute.

2. Another lesson learned requirement was that provisions to supply emergency electrical power to the presurizer heaters must be made without degrading the capacity, capability or reliablility of the onsite emergency power supply. The boards found the IMI-1 provisions to be acceptable in large part on the basis that two conditions would be satisfied: a) the heaters would not be connected until the load on the emergency power supply was stabilized; and b) that the undervoltage trip used to disconnect the heaters to protect the power supply frcrn a failure in the heater circuits would be set at a specified voltage and time delay. GPU has since changed the setpoints for toth the voltage and time delay on the basis that starting of motors on the emergency power supply would result in tripping th'e heaters. This indicates that the heaters will be connected without first stabilizing the load on the power supply. ,

The staff other than to classify it asisaaware of item."

" closed this situation but has taken no action

3. A principal purpose of the restart procteding was to decide, in view of the 1MI-2 accident, what new safety equiprJent was necessary and sufficient to protect the health and safety of the public during operation of TMI-1.

That process has been completely frustrated by the approval of technical specifications that permit operation of TMI-1 with equipment inoperable which the boards found was necessary to protect the public. For example, the following new safety equipment falls in this category: subcooling margin instrumentation, position indication for the pressurizer PORV and safety valves, emergency feedwater system flow instruments, and reactor coolant system high point vents.

UCS believes that the Comnission should take the following actions:

1. Direct the staff to identify any additional " lessons learned" require-ments or other items that the staff has certified to the Comssion as complete but which are no longer complete.

2. Direct the staff to identify any additional instances where the present status of TMI-1 does not conform with the GPU commitment. and/or board-imposed license condition that was relied upon in the adjudicatory decisions favorable to restart.

For items 1. and 2. above, the staff should also be directed to propose a i

method and schedule for resolving each safety issue identified, giving due consideration to the rights of the other parties.

3. Direct the staff to review the IMI-1 technical specifications and a:end them as necessary to require, as " limiting conditions for operation,"

the operability of all equipment relied upon by the boards as a basis for decision that IMI-1 can be operated without endangering the health and safety of the public. In addition, the staff should be directed to serve the parties with a complete set of the amended technical specifications and solicit the parties' coments on whether the amended technical specifications contain all such required limiting conditions for operation.

In sum, UCS believes that there are more items requiring resolution p' rior to a Commission vote on restart than were identified by the staff in its April 17, 1985 presentation to the Carnission. We anticipate the charge that these questions are being raised at the last minute. On the contrary, these are not

, new questions. They are issues which were purportedly settled during the restart proceedings but which have become unraveled by either inadvertent or intentional lack of attention to the conditions of restart imposed by or implicit in the ASLB and Appeal Board decisions.

j -

Sincerely, Robert D. Pollard Nuclear Safety Engineer

, s .

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, Ellyn . Weiss

General Counsel encl: Additional Safety Issues Requiring -

4 Resolution Prior To 1MI-1 Restart cc: TMI-1 Service List 4

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e ADDITICNAL SAFETY ISSUES RECUIRING RESOLUTION PRIOR TO THI-T RESTART i

j Union of Concerned Scientists DOCKETED 4 USNRC May 3, 1985

'85 HR-6 pjt :52

1. " Lessons Learned" Recuirements That Are Incomplete 1

a. NUREG-0737, Item II.D.1, Performance Testing of Boilih@g4faIEE.45dh..X9 j

Pressurized-Water Reactor Relief and Safety Valves (NUREG-0578, Section 2.1.2) i One of the lessons learned requirements was that reactor coolant system j

safety valves had to be . tested under the full range of expected operating con-ditions for design basis transients and accidents, which would include solid-1

. water and two-phase flow conditions. [Ref. 1.) As discussed below, data from testing Dresser Industry safety valves, including the specific model used as pressurizer safety valves at 7MI-1, showed that certain plant modifications and adjustments 1/ to the safety valves were necessary to prevent valve chatter 4 that could reduce relief capacity and damage the valve internals. GPU commit-ted to these changes, which were relied upon by the Appeal Board to resolve

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I j th2 issue.

Racent information indicates that GPU will not utilize the valve *

32ttings relied on by the Board. ,

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, ,1/ Tnese adjustments are called ring settings. Although there is some inter-action between the three separate ring settings, the principal effect of i

each ring setting is as follows: the lower ring setting affects the rate j

' at which the valve opens - and adjusts " accumulation," the pressure above !

the preset opening pressure at which the valve 'is fully open; the middle

! ring adjusts " blowdown," the pressure below the opening pressure at which the valve is fully reclosed; the upper ring setting is used to compensate for the effects of backpressure downstream of the valve. Accumulation and

' tiowdown are generally expressed as a percentage .of the pressure setpoint.

For example, a safety valve set to open at 2500 psi with 3% accumulation and 101 blowdown should open at 2500 psi, be fully open at 2575 psi and reclose at 2250 psi, (Ref. 2.)

.- l

.,..%-i,~, , n. -,,-,+~,-.-m- , -_y.- , , , _ - , - - -

1 1

, B&W provided the 'following evaluation of the results from the testing performed on the safety valves:

Recent review of the EPRI test data indicate that for certain combinations of lower and middle ring settings. the safety valves could chatter (rapid cycle valve opening and closing).

Two consequences of valve chattering have been noted during valve testing: a reduction in safety valve capability to relieve pressure while chattering and subsequent damage to valve internals. During 'he EPRI tests, this chattering phenomenon was noted onlv with valves with long inlet piping configurations. Howeve , present understanding of the effect ,

of ring settings on valve operation indicates that even with the valves mounted directly on the pressurizer nozzles, valve chatter could be postulated for a combination of ring settings significantly different from those used during the EPRI tests.

[Ref. 3.3 i

During the restart hearings conducted by the Licensing Board, UCS assert-ad, inter alia, that appropriate qualification testing of the safety valves had not been performed to verify the capability of these valves to function during normal, transient, and accident conditions. [Ref. 4.3 In resolving this question, the Licensing Board found that a valve testing program being conducted by the Electric Power Research Institute (EPRI) was adequate to reveal any design deficiencies in the safety valves used at THI-1. [Ref. 5.3 This subject was revisited during the reopened hearings held by the Appeal Board in March 1983. GPU testified that as a result of t.he EPRI tests

"[t]he safety valves were shown to be acceptable under all modes of operation after certain rodifications were made." [Ref. 6.3 The modifications which GPU testified were necessary in order to assure acceptable safety valve perform-cnce were to change "the safety valve inlet piping from a long inlet to a short inlet arrangement" [Ref. 7.3, and to readjust the safety valve ring set-

- tings to 'athe final EPRI ring settings which we used in the last five tests."

[Ref. 8.3 GPU's commitment to the latter modif t:stion has been abandoned.

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, During an NRC inspection of MI-1 conducted on January 21-24, 1985 (but not reported until March 18, 1985), the inspector noted that GPU had changed the safety valve ring settings. CPU furnished the inspector with documenta-tion from E&W dated in February and May of 1984 that purportedly supports the ravised ring settings. However, the inspector concluded that since the staff's safety evaluation report had been based on the earlier ring settings, there was a need to inform the Office of Nuclear Reactor Regulation (NRR) of the details. GPU representatives agreed "and committed to furnish the r-ing s2tting information to NRR by 3/31/85." [Ref. 9.] To date, UCS has not rseeived that info mation or any staff evaluation of it.

UCS believes that this safety issue must be resolved before restart. The safety significance of the ring settings for the pressurize safety valves is illustrated by the fact that after reviewing their safety valve ring settings

_ against the EPRI test results, Duke Power Company declared the safety valves inoperable and shut down Oconee Unit 2. [Ref. 10.3 The Oconee plants and IMI-1 utilize the same model of Dresser safety valves. [Ref. 11.] Further-msre, as discussed above, incorrect ring settings can result in inadequate relief capac' and da= age to the safety valves. Since the staff claims that the "limiti., nsient" for MI-1 is a suberitical rod withdrawal [Ref. 12.3 and such an accident could occur early in the restart process, resolution of this safety issue can not be postponed until after restart,

b. NUREG-0737. Item II.F.2, Attachment 1, Design and Qualification Criteria for Pressurized-Water Reactor Incore Thermocouples.

One of the " lessons learned" requirements was that additional instrumen-tation was needed for detection of inadequate core cooling. One requirement cpplicable to ir. core thermocouples that measure core exit temperatures was that "[t]he instrumentation should be environmentally qualified . . .. " [Ref.

13.3

-n-A GPU submittal to the staff dated May 20, 1983 stated that no extension ,

i of the March 31, 1985 ' deadline in 10 CFR 50.59 for environmental qualification '

of electrical equipment would be needed. In a submittal dated December 11, 1984, GPU " identified nine (9) items of equipment that were not qualified,"

t including the incore thermocouple instrumentation. [Ref. 14.3 On February 12, 1985, GPU applied for an extension of the deadline for qualifying the thermo-couples. GPU gave as its " reason for scheduler [ sic] extension" the following:

Qualification testing of the existing cables and connectors 4

is planned to be complete in April 1985 based on vendor bids.

The final report is due out in May 1985. Pending the outcome of the testing, replacement may be necessary. GPUN is pro -

ceeding in a parallel path with the testing to procure

! environmentally qualified replacement cables . . . .

[Ref. 15.3 -

b GPU offered as a "JIO" (Justification for Interim Operation)E the asser-tion that "[t]he safety function (of the incore thermocouple instrumentation 3 con be accomplished by alternate qualified instrumentation with existing i

procedural guidance." [Ref. 16.3 GPU asserted that the het leg resistance .

temperature detectors (RTDs) are the " alternate qualified instrumentation" to 4

b2 used in determining the subcooling margin of the reactor coolant system.

However, GPU also correctly informed the staff, albeit in a footnote, that

"[w]ith reactor coolant pumps running, the subcooling margin is calculated from the hot leg RTDs. If the RCPs are off, the subcooling margin is calcu-lated from the incore thermocouples." [Ref.17.3 i

i f

1/ Previously, such information has been termed a "JC0" (Justification for Continued Operation). The criteria for an acceptable justification for equipment not shown to .be qualified require, inter alia, either a) the equipment would perform its function before failing or b) there is other redundant, qualified safety-grade equipment available to perform the same '

safety function.

The staff appears to have overlooked GPU's footnote. In a letter to GPU dated March 29, 1985,1' the staff stated that "[w]e have reviewed the JIOs and l

conclude that they are sufficient to support operation during the requested '

extension period," and granted GPU's requested extension of the March 31, 1985 environmental qualification deadline. [Ref. 18.3 However, even assuming, arguendo, that the instrumentation used to datermine subcooling margin from the hot led RTDs is environmentally quali-fied, that intrumentation can not substitute for the incere the-mocoup5es b2cause it can not be used unless the reactor coolant pumps are running.

Under design basis accident assumptions, the reactor coolant pumps will not be running. The Commission's regulations (GDC-17) require postulating the loss of offsite eler ical power, which is the only source of power for the pumps.

Furthermore, even if offsite power is not lost, the INI-1 reactor operators hcve been trained and instructed to shut off the reactor pumps in certain accident situations. Thu s , UCS concludes that NUREG-0737 Item II.F.2 is not co=plete because the incore thermocouples are not envirommentally qualified. .

UCS also concludes, contrary to the staff, that GPU has not provided a valid justification for operation of TMI-1 without environmentally-qualified incore thermocouple instrumentation. One or the other must be done before operation of THI-1 can go forward uunder the rules that apply to all plants.

2. GPU Commitments or License Conditions That Are Not Met One of the lessons learned requirements was that provisions to supply cmergency electrical power to the pressurizer heaters must be made without d2 grading the capacity, capability or reliability- of the onsite emergency

}/ UCS received a copy of this letter at 4:00 p.m., Friday, April 15, 1985.

f .

, power supply. During the restart proceeding, UCS contended that the provi-sions at IMI-1 to protect against the loss of the emergency electrical power supply as a result of an electrical fault (e.g., short circuit) in the non-safety grade pressurizer heaters or their circuits were inadequate. [Ref.19.)

This safety issue was resolved by the Boards relying upon, inter alia, GPU's commitments that two conditions would be satisfied at TMI-1. First, the Boards relied upon GPU's testimony that "[t]he undervoltage relays will trip open the [ pressurizer heater circuit] breaker if the voltage on the ES

[ engineered safeguards] bus drops to 430v for 1.5 seconds." [Ref. 20.]

Second, the Boards relied upon GPU's testimony that the pressurizer heaters will not be connected to the emergency power supply unless " adequate diesel generator capacity is available and all systems have stabilized." [Ref. 21.]

GPU apparently has no intention of adhering to these two conditions.

During routine safety inspections of IMI-1 conducted on January 3 -

February 3, 1984 NRc inspectors noted that GPU had changed both the voltage end time delay setpoints for the undervoltage trip. The inspectors reported -

that the setpoints were changed from those relied upon by the Appeal Board to avoid tripping of the pressurizer heater circuit breakers "due to motors starting on the class 1E bus." (If motors are being started, then the load on the power supply has not been stabilized.) Although aware that the Appeal Board decision was based on different setpoints, the inspection report states that "[t]his item is closed." The relevant portion of the inspection report follows:

The setting specified in the Licensee [ sic] Bnard Hearing ALAB-729 was 430 volts at 1.5 seconds. The setting by the licensee is 420 volts at 1 second.

The ALAB-729 relay setting of 430 volts would protect the class 1E system. However, the purpose of CLI-79-8 would not be achieved due to undesired tripping. This setting is 907. of normal (voltage) which would be reached due to motors starting

s

, on the class 1E bus. The setting of 420 volts is 85% of l no mal which allows 80% for motor starter coil pickup plus 55 l voltage drop through the control transformer and control I wires. The time setting has been reduced from the ALAB-729 i j requirement of 1.5 seconds to 1 second. The 855 voltage

setting at 1 second assures the primary objective of pro-

tecting the _ class 1E system. This item is closed.-[Ref. 22.]

l_ It is not clear what evaluation, if any, was done by the staff to support 4

the conclusion that revised undervoltage trip setpoints achieve "the primary

.i objective" of protecting the onsite emergency power supply. It is clear, however, that the revised setpoints are not equivalent to those relied upon by the Appeal Board. For example, an electrical fault that caused the voltage to 1

drop to between 420 and 430 volts would not result in tripping of the circuit 4

4 breakers no matter how long the fault persisted if the unauthorized setpoints are used. In contrast, the same fault would result in tripping the breaker if j the setpoints relied upon by the Appeal Board are used. Ihus, to some degree, j the revised setpoints degrade the capability and reliability of the emergency power supply. In any event, the staff and GPU can not unilaterally change J .

ccm: nitments relied upon by the Boards.

1

+

UCS considers the abandonment of the commitment not to connect the pres-surizer heaters to the emergency power supply unless the loads have stabilized i ~

to have even more safety significance because a failure of the emergency power system at this stage could imperilall of the plant's safety systems. Further-

~

Dere, if that commitment had not been abandoned, there would ' apparently have i bsen no need to depart from the specific voltage and time delay setpoints for the undervoltage trip which were relied upon by the Appeal Board.

3. TMI-1 Technical Specifications That Negate Board Decisions A principal purpose of the restart proceeding was to decide, in view of I the THI-2 accident, what new safety equipment was necessary and sufficient to l

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i l protect the health and safety of the public during operation of TMI-1 That l process has been completely frustrated by the approval of license technical i

specifications that permit operation of INI-1 with equipment inoperable, even though the Eoards found that equipment to be necessary to protect the public.

Examples of instances where this has occurred are as follows:

a. Subcooling Margin Instrumentation. A lessons learned requirement was that primary coolant saturation meters should be installed to provide online indication of coolant saturation condition. (Ref. 23.3 GPU testified that j this requirement was met by the installation two "subcooling margin monitors,"

i one for each loop. However, the technical specification permit unrestricted J

j operation of INI-1 with one ~subecoling meter inoperable all the time and permit l operation for an additional week after failure of the second. [Ref. 24.3 i

!_ The purported basis for this technical specification is that "[a)1 ternate i

! indications are available for Saturation Margin Monitors using hand calcula, i tion .f. 25.] However, the use of " hand calculations," which were avail-able during the TMI-2 accident, obviously does not meet the lessons learned l requirement to install instrumentation "to provide on-line indication of i

i coolant saturationcondition." [Ref. 26.] Furthermore, the only temperature indications available for such hand calculations, other than the hot leg l resistance temperature detectors which provide the temperature input to the subcooling meters, are the incore thermocouples. As discussed above, the l staff has p; oposed allowing INI-1 to operate without environmentally qualified incore thermocouple instrumentation on the grounds that the incore j

i thermocouples are backed up by the subcooling meters. In any event, the l

! lessons learned requirements specified that both the incore thermocouples and i

saturation meters must be provided and GPU tesimony during the restart i

proceeding was"that both would be.

E l

b. Direct Indication of PORV and Safety Valve Position. A lessons Isarned requirement was that reactor system relief and safety valves shall be provided with reliable positive indication of their position in the control room. [Ref. 27.] GPU testified that this requirment had been met by the installation of redundant position indicators for the PORV and by a single position indicator for each safety valve. Specifically, discharge flow from the two pressurizer code safety valves and the PORV is measured by differen-

~

tial pressure transmitters connected across elbow taps downstream of each volve. In addition, an acoustic monitor is provided to detect flow in the PORV discharge line.

~

However, the technical specifications permit operation of THI-1 with cither one of the PORV instruments continuously inoperable and permit con-tinued operation for a week following the failure of the other PORV instrument and both safety position indications. [Ref. 28.] The basis purported to justify this relaxation of the lesson learned requirements is that alternate indication of the position of the safety valves and the PORY is available from the thermocouples in the valves' discha ge pipes. It is precisely these dswnstream thermocouples that proved inadequate during the TMI-2 accident and Icd to the requirement for reliable position indication.

c. Emergency Feedwater System Flow Indication. A lessons learned rsquirement was that "(s)afety-grade indication of auxiliary feedwater flow to i

occh steam generator shall be provided in the control room." [Ref. 29.3 GPU testified that this requirement was ret by the installation of redundant EFW flow instruments for each steam generator.

. However, the technical specifications permit operation of TMI-1 with one flow instrument for each steam generator inoperable all the time and permit cantinued operation for a week following failure of the other EFW flow instru-

1 l

1

)

~

, m2nts. [Ref. 30.3 Allowing operation with one flow instrument inoperable does i

n3t meet the requirement for si.fety-grade instrumentation because the single fcilure criterion is not met with only one flow indicator per steam generator.

The availability of local EW flow indication does not satisfy the requirement that the indication be in the control room,

d. Reactor Coolant System High Point Vents. A lessons learned require-mant was that high points vents be installed inorder to vent non-condensible gases from the reactor coolant system which could otherwise block natural circulation cooling of the core. The lessons learned requirements applicable to the high point vents specified that although each vent path need not meet the single failure criterion, a " degree of redundancy should be provided by powering different vents from different emergency buses." [Ref. 31.]

The evidentiary record indicates that TMI-1 complied with this require-e2nt by installing four vents - one on each loop, one on the pressurizer and

, one on the reactor vessel. The design is such that the vents on the vessel, pressurizer, and loop B are powered from one electric power supply and the vant on loop A is powered from a redundant power supply.

However, the technical specifications permit operation of THI-1 with one Icop vent inoperable all the time and permit continued operation for 30 days following failure of another vent. [Ref. 32.] Thus, if TMI-1 is operating t

with the loop A vent inoperable, a single failure of the power supply for the pressur'izer, vessel and ' loop B vents would render all high point vents inoper-cble. The technical specifications therefore violate the lesson learned rGquirement that the vents collectively should meet the single failure criter-ion. Furthermore, since the vents on the pressurizer and the vessel can not remove gas from the high point of the loops, the same situation (i.e. , the inability to restore natural circulation) would prevail if THI-1 were in oper-

. . - . -._y~.. _.y .

ation with the loop B vent inoperable. That is, an accident accompanied by a single failure of the power supply for the loop A vent would leave TMI-1 with no operable loop vent. The basis for this technical specification provides no justication either for allowing operation with one loop vent continuously inoperable, or for 30-days continued operation with both loop vents inoperable.

UCS concludes that the TMI-1 Technical specifications must be amended to require, as " limiting conditions for operation," the operability of all equip-cent relied upon by the Boards as a basis for the decision that IMI-1 can be operated without endangering the health and safety of the public. In addition, we believe that another requirement should be added to the technical specifications.

The technical specifications permit continued operation for some period of time following failure of safety equipment that is subject to a limiting condition for operation. If repairs can not be made with the specified time, the technical specification require that the plant be shut down. UCS believes .

I

that GPU should be required to assess, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after detecting the failure, whether the equipment can be restored to an operable status within the time continued operation is permitted without that equipment. When it -is concluded that repairs can not be accomplished within the allowable time a ,

p2riod, the plant should be immediately shut down. There is no justification for continued operation, even for a limited time, if it is known that the equipment can not be repaired prior to the time the plant will be required to ,

shut down.

4. References. See next page. l l

e

_.. _ _ . . _ _ _ .~

c References I

1. NUREG-0737. Item II.D.1. See also, NUREG-0578, Section 2.1.2.

2. Appeal Tr. pp. 151-152 (Lanese), March 7, 1983. -

3. J. H. Taylor, Manager, Licensing Services, B&W, letter to R. C. DeYoung, Director, IE,-NRC, October 15, 1982, p. 1.

4 See LBP-81-59, 14 NRC 1211, 1375 (1981).

5. 14 NRC 1378-79. .

6. Jones and Lanese, fol. Appeal Tr.111 at 3-4, March 7,1983, emphasis added.

7. 16., at 4

8. Appeal Tr. 423 (Correa), March 16, 1963, emphasis added.

9. NRC Inspection Report No. 50-289/85-03, March 18,1985, p. 3.

10. Frank Miraglia and Gus Lainas, Division of Licensing, NRC, memorandum to James Knight, Division of Engineering, NRC, " Review of Licensee / Vendor Letter on Dresser Safety Valves," November 3,1982.

11. Thomas Ippolito, Division of Licensing, NRC, memorandum to Dennis Crutchfie*d, Division of Licensing, NRC, " Review of Licensee / Vendor Letters on Dresser Safety Valves," December 30, 1982, enclosure, Table 1. .

12. Id.

13. NUREG-0737, Item II.F.2, Attachment 1, p. 3-118.

-14 Harold Denton, Director, NRR, NRC, letter to Henry Hukill, Vice President and Director - TMI-1,,GPUN, March 29, 1985, p. 1.

15. GPUN letter no. 5211-85-2018 to Harold Denton, NRC, February 12, 1985,

p. 1, emphasis added.

16. Jd., enclosure,p.,6.

17. Jd., e4 closure, p. 3, n. 1.

18. Harold Denton, Director, NRR, NRC, letter to Henry Hukill, Vice President and Director - TMI-1, GPUN, March 29, 1985, p. 2.

~

19. See, e.g., ALAB-729,17 NRC 814, 855-859 (1983).

20. 17 NRC 858, footnote omitted.

21. 17 NRC 860, footnote omitted, emphasis added.

l

r- .

. . 22.

NRC Inspection Report No. 50-289/84-01, March 17, 1984 (NRC transmittal 1etter dated April 9, 1984), p. 5.

23. NUREG-0578, pp. A-11, A-12. .

24 Technical Specification 3.5.5 ( Amendment No.100) and Table 3.5-2

( Amendment No. 78).

25. Technical Specification 3.5.5. p. 3-40b ( Amendment No.100).

26. NUREG-0578, - p. A-12.

27. NUREG-0578, pp. A-9, A-10.

28.

Technical Specification 3.5.5 (Amendment No. 100) and Table 3.5-2 -

(Amendment No. 78).

29. NUREG-0578, p. A-32.

30. Technical Specification 3.5.5 ( Amendment Ft . 100) and Table 3.5-2 (Amendment No. 78).

31 NUREG-0737 Item II.B.17 Reactor Coolant System Vents.

32. Technical Specification 3.1.13 ( Amendment No. 97).

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