ML19305C882
| ML19305C882 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 03/17/1980 |
| From: | Weiss E SHELDON, HARMON & WEISS, UNION OF CONCERNED SCIENTISTS |
| To: | METROPOLITAN EDISON CO. |
| References | |
| NUDOCS 8004100562 | |
| Download: ML19305C882 (28) | |
Text
9 ECL.i!ID CORRESPONDENCE UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
)
In the Matter of )
)
METROPOLITAN EDISON COMPANY , )
et al., ) Docket No. 50-289
) (Restart) dD ' I I/
(T hre e Mile Island Nuclear Station ) fS3 4 Unit No. 1) ) G \
f,
)
UNION OF CONCERNED SCIENTISTS ' b h*.l ' 13 '.!22 >
ANSWERS TO LICENSEES FIRST i ,,7., ng g3 7 :.tr!
SET OF INTERROGATORIES \go --'E.'l.1,q
'; 3rd 3 fC. "WA Y \j$
Question 1-1 g Explain how the TMI-2 accident demonstrated that reliance on natural circulation to remove decay heat is inadequate.
Answer 1-1 Prior to- the operator shutting off- the reactor coolant i pum ps , an attempt was made to use. forced cooling. After the last reactor coolant pumps were tripped, a rapid heat-up of the core commenced. This demons trated that natural circulation was in adequ a te . Later in the accident sequence, forced cooling was re-established by starting a reactor coolant pwnp.
Question 1-2 Does UCS contend that reliance on natural circulation to remove decay heat was inadequate at all stages in the sequence of events at TMI-2 following the trip of the feedwater system?
If so, explain the basis for the contention. If not, identify the point in the sequence of events when UCS contends that 8 004100 IITd2--
i reliance on natural circulation to remove decay heat became in adequat e .
i Answer 1-2 I
- For all stagra in the accident sequence at which an attempt h i
was made to rely on natural circulation it was inadequate. Our basis for this statement is that it was not until re-start of a reacto coolant ptmp that adequate core cooling was provided.
Question 1-3 Does UCS contend that natural circulation is (a) impossible or (b) ineffective to remove decay heat after voids have formed in the primary cooling system? I f so , provide the technical bas is for the contention and reference any documents which in i
UCS' view support the contention.
Answer 1-3 Our contention is that the accident at TMI-2 demonstrated i th a t the formation of voids can render natural circulation 4
i ineffective to remove decay heat. If the interrogatory is direc ted towards a theoretical question, the answer to whether void formation makes natural circulation " impossible" or "
ineffec-tive" depends on the volume of the void and other variables such as the extent of flow blockage caused by physical damage to the core. Reference NUREG-0 5 78 , particularly p. A-2, the paragraph beginning at line 11.
! Question 1-4 ,
What is the basis for the statement that during the TMI-2 accident it was "necessary" to opera te at least one reactor 4
coolant pump.to provide forced. cooling of the core?. Explain '
-o why, even assuming forced cooling to be required, operation i of the high pressure injection system would not have sufficed to cool the core. ,
Answer 1-4 The basis is that during the accident sequence, a number of efforts were made to bring the accident under control. All such efforts were ineffective and the cora was not adequately l t cooled until forced' cooling was established.
In the actual accident as it occurred, both the high pre ss ure injection system and the reactor coolant pump were us ed to re-establish adequate core cooling. It is speculative (in the absence of tests and inspection of the TMI-2 core )
whether the extent of physical damage to the core may have precluded the effectiveness of the high pressure injection sys tem alone. In addition, as the staff has noted on p. A-2 of NUREG-05 78, use of the high pressure ECCS for this purpose "may exceed the previously unders tood and accepted design l
basis."
Moreover, UCS no te s that this interrogatory appears to be ;
premised on the assumption that the time at which the high pressure injection system is actuated is irrelevant. UCS disputes this premise.
Question 2-1 Explain what you mean by operation of the emergency core .
cooling sys tem in a " bleed and feed" mode. Describe all flow paths associated with such mode of operation.
e Answer 2-1 The " bleed and feed" mode referred to in the contention was a result of the NRC staff 's statement that "the feed and bleed mode of reactor coolant system operation can be used to remove decay heat from the reactor." NUREG-0578, P. A-1.
UCS does not know the details of the staff's basis for this e
statement, nor do we understand the flow path envisioned by th e s ta f f . It was included in the contention because we under s tood the staff in the above-quoted language to be suggesting that such a mode could provide adequate core cool-ing.
Question 3-1 Does UCS contend that natural circulation cannot be main-tained at hot stand-by conditions without use of pressur-izer heaters and controls ? I f so , explain fully the basis for <
the contention.
Answer 3-1 e The answer depends on the circumstances involved. For ex ample , when the pressurizer is only partially filed with !
water, use of the pressurizer heaters, pressurizer spray and the pressurizer relief and safety valves or some combination thereof are needed to control pressure. Control of pressure is essential to ensure that natural circulation can be esta-blished. Other means of controlling pressure, for example, -
high pressure injection makeup and letdown, are most effec-I tive when the pressurizer (and, the entire reactor coolant i
l
sys te m , including the reactor vessel) are full of water.
The time required to fill the pressurizer (or the entire reactor coolant sys te m ) and the difficulty of maintaining control of pressure in a water-solid system makes the use of pressurizer heaters and controls essential in some circum -
stances. UCS contends that heaters and controls are necessary to give reasonable assurance that natural circula-tion can be maintained.
Question 3-2 !
Explain the basis and provide document references for the ,
statement that the staff recognizes that pressurizer heaters j and associated controls are necessary to maintain natural cir-
- culation at hot s tand-by conditions.
Answer 3-2 i
NUREG-0 5 78, p. A-1: " Natural circulation cooling of the primary"sys tem requires the use of the~ pressurizer to maintain a suitable over-pressure on the reactor coolant sys tem. "
Question 4-1 Refer to the information contained in Section 2.1.1.3.1 of the Restart Report. Does UCS contend in the light of that information that adding pressurizer heaters to the on-site emergency power supplies will degrade the capacity, capability and/or reliability of these power supplies? If so, explain fully the basis for the contention. .
Answer 4-1 Th material read was S2.1.3.1, page 2.1-5,_2.1-Sa, ,
2.1-6 all in Amendment 3 and page 2.1-7, Amendment 6. In UCS 's view , that information confirms that adding the pressurizer
heaters , which are not safety-grade , to the on-site emergency power supplies will degrade the capacity capability and/or reliability of these power supplies.
Section 2.1.1.3.1.2 contains a s tatement alleging that the separation and isolation of safety equipment from non-safety equipment will be in accordance with Regulatory Guide ,
1.75. However, the material immediately following that statement demons tra tes that Regulatory Guide 1.75 will not be met. S pecifically , the use of the undervoltage relay to de tec t a fault is prohibited by Reg. Guide 1.75, which states tha t it is prudent to preclude the use of interrupting devices actuated only by fault current as accepta-ble devices for isolating non-Class IE circuits from Class IE circuits or associated circuits.1/
The Regulatory Guide explains that neither a trip signal derived f rom f aul t current or its '~ef fect's is acceptable.
Undervoltage is an ef fect of the fault current.
In addition, Sec ti on 2.1.1.3.1.5 represents a violation of the single f ailure criterion. It s ta tes that this design can "at most, cause the loss of one 480 volt ES [ engineered safeguards] sys tem. " The premise is that such a design is acceptable. However, in analyzing the adequacy of this design, one must assume the occurrence of another single failure in i
~
addition to loss of offsite power; GDC 17 requires that the onsite power system be capable of perf orming its safety function i
I 1/ Position Col . Basis.
l i
_7_
af ter a loss of off site power plus a single failure in the onsite emergency power sys tem. This design f ails to meet that cri terion . If one emergency diesel fails to start and a fault oc curs in the non Class IE pressurizer heater circuits , the ne t result would be loss of both engineered safeguards power supplies. Since the pressurizer heaters are not Class IE, pos tulating a circuit fault in the pressurizer heaters in addi-tion to f ailure of one diesel generator is consistent with the single failure criterion.
Furthermore, reliance on operator action to connect the heaters to the ES power supply and for load shedding is a vio-lation of 10 CFR 50.55 ( a )( h ) , which incorporates IEEE 279.
IEEE 279 requires that actions needed to protect the health and safety of the public should be initiated automatically.
This list is not exhaustive because UCS has not yet completed its analysis and will not do so until all amendments to the Restart Report have been submitted.
Ques tion 4-2 Refer to the safety evaluation performed by the NRC Staf f at pages C8-3 through C8-8 of the January 11, 1980 Status Report on the Evaluation of Licensee's Complaince with the NRC Order Dated August 9, 1979. Does UCS contend in the light of that information that adding pressurizer heaters to the onsite emergency power supplies will degrade capacity, capability and/or reliability of these power supplies? I f so , explain fully the basis for the contention', indicating with particularity the inadequacies of the NRC Staff safety evaluation.
Answer 4-2 Yes, UCS's contention stands in the light of that information.
The basis for the contention is the same as discussed in the answer to question 4-1. Furthermore, the staff's " Clarification" items Nos. 3, 4, 5 and 7 on page C8-6 appear to be designed to allow approval of the design proposed in the Restart Report rather than for the purpose of clarifying existing regulations or providing adequate protection f or the public.
S pecifically , Clarification items Nos . 4, 5 and 6 ! violates the Commission's regula tions and Regulatory Guide 1. 75. See the discussion in the answer to 4-1 above.
Question 4-3 Explain in detail in what respect UCS contends that GDC 17 would be violated.
Answer 4-3 See response to 4-1, above.
Question 5-1 Provide the technical basis for UCS' contention that proper operation of power operated relief valves, associated block :
valves and the instruments and controls for these valves is esential to mitigate the consequences of accidents.
f 2/ No. 6 does not itself violate Regulatory Guide 1. 75. However, the staff's finding that the decign is acceptable is not based 'on conformance with the Reg. Guide.
I-l
f i Answer 5-1 The staff noted in NUREG-05 78, p. A-6, that the "[p3 roper opera tion of reactor coolant system relief and safety valves is vital for conformance to [GDC 14, 15 and 30]." The basis for this statement is contained in the discussion in NUREG-05 78 which follows the quoted statement. A notable omission from the staff's discussion is the role of the relief valves in mitigating ATWS events. The staff acknowledges on the last paragraph of page A-7 that it would be " prudent" to consider this, but that it has not done so.
Que s tio n 5-2 Describe the accidents or categories of accidents to which UCS contends that the first sentence of Contention No. 5 applies.
Ans wer 5-2 The types of accidents during which proper operation of the relief valve is essential include the accident at TMI-2 on March 28, 1979 and the one at Crystal River on February 26, 1980. It would include any type of accident which would cause the relief valve to open.
UCS does not contend here that there are accidents during which the relief valve is required to open for mitigation pur-pos es . We do contend that these are accidents'which caused the relief valve to open and that proper operation of the PORV, associated block valve and their instruments and controls is -
necessary to terminate the subsequent release of coolant water-4 Question 5-3 Does UCS contend that .satisf actory mitigation of the acci-den ts identified in answer to interrogatory 5-2 is not possible
without the opening or manipulation of the power-operated relief valve? If so, explain fully the basis for the contention.
- Answer 5-3 As discussed in the answer to 5-2 above, UCS's contention focusses on terminating the coolant release, not the necessity of opening the PORV.
Question 6-1 l
Identify the appropriate qualification testing which UCS contends has not been done and should be done to verify the capability of the reactor coolant system relief and safety ;
valves to function during normal, transient and accident condi-tio ns .
Answer 6-1 The testing necessary is that described on pages A-6, A-7 and A-8 of NUREG-05 78. UCS 's tes timony will consis t of an eva-luation of the propos ed qualification testing program and the tes t results.
Question 7-1 Describe the instrumentation proposed by UCS for direct measurement of the water level in the fuel assemblies. Explain whether, and if so how, such instrumentation would provide relia-4 ble and unambiguous it.dication of adequate core cooling under all accident conditions.
Answer 7-1 ,
UCS contends that the provision of water level ins trumen ta-
- _ tion is required by the regulations . We are aware of no techni- ,
l cal reason which would preclude the design and installation of I
i 1 e l
of such ins trumentation. Water level measurements are typically provided for steam generators, pressurizers and boiling water reactor vessels. It is also our information that the ACRS has recommended that Met Ed install coolant-level instrumentation as a condition for TMI-l restart, and that Met Ed is fighting this. A copy of the ACRS letter is attached.
We assume the license is not asking UCS to propose a design for the instrumentation. '
Question 9-1 Provide a technical basis and explanation for UCS ' contention that the TMI-2 accident was subs tantially aggravated by the closing of two auxiliary feedwater system valves.
Answer 9-1 The original basis for the UCS contention can be found in Sec tion 2.1. 7,2 . 2. 3 and the corresponding portions of Appendix A, NUREG-05 78 .
Investigations since the time that this contention was written have indicated that the auxiliary feedwater valves were re-opened quickly enough so that their initial closed positions appears not to have had a significant impact on the further course of the accident. However, it is clear that total loss of feedwater for a prolonged period of time would have seriously aggravated the accident. Moreover, the fact i
that the auxiliary feedwa ter sys tem was totally incapacitated .
aggravated the accident by distracting the operators. In any event, the basis for UCS's contention is that the design of the auxiliary feedwater system and other systems at TMI-1 does not
i l
mee t the single failure criterion.
Question 9-2 Explain the basis for any disagreement which UCS may have 4
with the conclusion of the President's Commission (page 94 of its i report) that '"[ t]he loss of emergency feedwater for 8 minutes had i
no significant effect on the outcome of the accident."
Answer 9-2 At this ; UCS has no particular disagreement with the
, statement, although we have not reviewed in detail the background i
! documents which may support the conclusion. We note, in addition, our agreement with the next sentence of the President 's Commission a
i re po r t . "But it did add to the confusion that distracted the 2
operators as they sought to understand the cause of their primary
, problem." ,
Question 10-1 1
Identify each safety function related to the core cooling ,
i and containment isolation systems as to which UCS contends that I 1 the design of a safety system must be modified so that no opera-tor action can prevent the completion of such safety function once initiated. '
Answe r 10-1 UCS has not yet identified all systems other than ' ECCS which I
mus t be modified. The way in which such systems can and will be identified is by reviewing the descriptive information in the SER i and the PSAR pertaining to the design of the protection sys tem and accident an alys es . This will identify the safety functions per-t i
formed by the protection system for which credit is taken in the accident analyses. UCS's contention is that the design of all TMI protection systems mus t be modified so that no operator action can prevent the completion of a safety func-tion once initiated. It is UCS 's understanding that all B&W protection systems are characterized by the same deficiencies as is the ECCS.
Question 10-2 Explain as to each safety function identified in answer to interrogatory 10-1 the criteria proposed by UCS for determin-ing completion of the safety function.
Answer 10-2 The criterion proposed by UCS for determining completion of the safety functions (identified by the method outlined in 10-1, above) is that no assumptions in the safety analysis shail be 4
viola ted . Tha t is , the safety systems should be designed such I
i that operator action does not prevent the accomplishment of the safety functions assumed in the safety analysis.
Question 10-3 Describe as to each safety function identified in answer to interrogatory 10-1 the design modification (s) proposed by UCS_so that no operator action can prevent the completion of such safety function once initiated.
Answer 10-3 j We do not propose any particular design (s) to achieve the goal outlined above. The approach to the development of any particular
design could be similar to the design of operating bypasses (See Section 4.12 of IEEE 279). For example, as to the NRC staff's post-TMI requirement that ECCS be operated for some ,
minimum time or until some temperature specification is met, the switches used by the operator to stop the ECCS pumps or r
to close injection valves could be interlocked to incapacitate i
those switches until the time or temperature limits are satis-fled.
Question 12-1 Identify each item of failed equipment in the containment and auxiliary buildings which is covered by Contention No. 12.
Answer 12-1 UCS has asked the staff 'to identify each piece of such equipment which f ailed and the way in which it failed. (See j UCS Interrogatories to the NRC Staf f , Nc. 21 ) As yet we have had no response. There f ore , we are not yet able to answer this question.
Question 12-2 State as to each item of equipment identified in answer to interrogatory 12-1 whether UCS contends that the equipment failed because of the TMI-2 accident environment and, if so, the
. technical basis for such contention.
! Answer 12-2 See the answer to 12-1 above.
i .
l Question 12-3 r
l Provide as to each item of equipment identified in answer l
l to interrogatory 12-1 the basis for UCS' contention that such
equipment was previously deemed to be environmentally qualified.
Answer 12-3 See 12-1, above. The basis for UCS 's contention that this equipment was previously deemed to be qualified was the (perhaps
, erroneous ) assumption that TMI-2 met the NRC regulations at the time of licensing. That is, that equipment needed for safety functions, whether automatically or manually initiated, complied with GDC 4.
Question 12-4 Identify each item of " safety-related equipment" in the
, TMI-l containment and auxiliary buildings which UCS contends
+
mus t be demonstrated to be environmentally qualified.
Answer 12-4 We cannot identify the specific safety equipment at this time.
The criterion for identifying such equipment which UCS advocates is "s truc tu re s , sys tems and components important to safety. . .
(GDC 4) We have asked the staff to inform us of what equipment in TMI-l is covered by GDC 4. (See UCS Interrogatories to the NRC staf f, No. 113). As yet, we have no answer.
! Question 13-1 Describe the accidents which UCS contends are credible and not bounded by the TMI-l design basis accidents.
Answer 13-1 l All accidents with consequences beyond the P ar t 100 limits-and/or sequer.ces of events previously considered so improbable as to be incredible are not within the design basis. As we r -.,-. - - . , . , y -,,,-g - .
have previously explained, it is our view that the NRC has no technical basis which would justify excluding possible accident sequences on the basis of probability. The following accident sequences identified in WASH-14 00 are examples of scenarios with potentially catastrophic consequences which should be considered in the licensing process: PWR 5, PWR 4, PWR 2. They are credi-ble and are not bounded by the TMI-l design basis accidents.
Question 13-2 Explain as to each accident identified in answer to interro-gatory 13-1 the nexus between such accident and the TMI-2 accident.
Answer 13-2 It is UCS's contention that the TMI-2 accident, which involved a combination of design errors, equipment failures and human errors significantly beyond what NRC had previously deemed probable has established that the method by which the staff classifies possible accidents as either within or outside the group of " credible "
accidents is fundamentally faulty. Our view is supported by
- the conclusions of both the short and long term TMI Lessons Learned Reports. ( See NUREG-05 78, p. 16-17; NUREG-05 85, p.
3-5 . ) Both call for a thorough reconsideration of the current concept of design basis events, because TMI exceeded the current design basis and was a significant precursor of a core-melt.
Thus, the nexus to the accident is clear: it was the accident that demonstrated the inadequacy of the current staff practice of excluding major reactor accidents from consideration in the licensing process. The specific PWR
_17-sequences are given as examples of such serious accidents which cannot be shown to be so improbable as to be incredible and the consequences of which have not been considered for TMI-1.
Question 13-3 j Explain what UCS means when it contends that an accident is not " bounded" by the design basis accidents for TMI. Indicate ,
in particular as to each accident identified in answer to interro-gatory 13-1 whether the term " bounded" refers to accident events or accident consequences or both.
Answer 13-3
" Bounded" refers both to accident events and accident co~
sequences.
Quest on 13-4 Describe the criteria proposed by UCS to be used for select-ing credible accidents to be considered. :
Answer 13-4 All possible accidents with potential consequences which would exceed the limits contained in 10 CFR Part 100 should be con-sidered. UCS could not support any method of selecting acci-dents based on probability of occurrence because there is no ,
known method of determining the actual probability of any accident sequence with sufficiently narrow error bands to justify the present "all-or-nothing " approach.
Question 14-1
, Define " adverse effect on the integrity of the core." j Answer 14d The ECCS criteria in 10 CFR 's 50.46 establish a set of
t conditions which must not be exceeded. In DCS's view, an
" adverse effect on the integrity of the core" would occur if any of those limits were exceeded for any reason, not just i malfunction of ECCS. These include peak cladding temperature, maximum cladding oxidation and embrittlement, maximum hydrogen generation and coolable core geometry, which involves rod swelling and rupture.
Question 14-2 Identify each system or component relating to the core cooling system presently classified as non-safety-related which UCS contends must be identified and classified as compo-nents important to safety and required to meet all safety-grade design criteria.
, Answer 14-2 See the UCS answer to NRC Staff Interrogatory No. 14-1.
Question 14-3 i 1 Explain how each system or component identified in answer to interrogatory 14-2 can directly or indirectly affect tempera-tu re , pressure, flow and/or reactivity. !
I Answer 14-3 The following can affect pressure: pressurizer heaters ,
pressurizer relief valves and associated block valve, pressurizer safety valve, chemical and volume control systems.
The following can affect temperature: reactor coolant pumps, auxiliary fe gdwater , main feedwater and condenser bypass systems.
Reactor coolant pumps affect flow.
l l
l
4 The following can affect reactivity: boric acid transfer systems, chemical and volume control systems and all items listed above which can affect temperature.
In UCS 's opinion , it is obvious how each of the above affects the parameter indicated. Of course, items listed as affecting
! pressure can also indirectly affect flow, temperature and reacti-vity. Specifically, f ailure of any of 'these items which results in a decreasing pressure could result in voids, thereby reducing flow, increasing temperature, and decreasing reactivity.
Question 14-4 Explain how each effect identified in answer to interrogatory 14-3 can have an adverse effect on the integrity of the core.
Answer 14-4 -
Maintaining the integrity of the core depends upon keeping these parameters - temperature, pressure, flow and reactivity -
within allowable limits. Recognition of this face was the basis for developing the limits in the ECCS criteria. Obviously, any effect which results in exceeding the limits set by the regula- ;
tion is, by definition, an adverse effect on the intergrity of the core.
Question 21 -1 (a-d) j With respect to each individual whom UCS intends to call as a witness in this proceeding:
- a. Identify by name, address and affiliation each such ,
individual;
- b. State the educational and professional background of each individual, including occupation and institutional affilia'-
tions, publications and papers;.
- c. Identify the contention as to which each such individual will testify;
- d. Describe the nature of the testimony which will be presented by each such individual, including an identification of all documents which the individual will rely upon in the te s timony .
Answer 21-1 (a-d)
The answers to these questions have previously been provided by UCS in response to NRC Staff Interrogatories Nos . 1 and 2, copies of which were served on the licensee. (January 18, 1980)
Question 21-1 (e)
Identify by court, agency or other oody, proceeding, date and subject matter all prior testimony by each such individual.
Ans wer 21 - 1 (e)
Mr. Pollard has presented sworn testimony in-the following cases:
- l. New York Public Service Canmission Case No.
80003, Re: J amespor t Nuclear Power Station, Profiled testimony was filed on or before March 18 , 1977. Supplemental testimony was provided in July, 1979. The general subject matter was the safety hazards and economic uncertainties inherent in the Jamesport project.
- 2. Texas Utilities Generating Co. (Comanche Peak Nos . 1 and 2), Docket Nos 50-445, 50-446.
l Mr. Pollard, as AEC Project Manager, sponsored Amendment No. 1 to the Staff Safety Evaluation Report, in November or December, 1974. The subject areas included conformance with ECCS I
Criteria and GDC 4. .
Mr. Pollard presented limited appearances in the following l
proceedings :
- 1. Virginia Electric and Power Co. (North Anna 1 and 2), Docket Nos. OL-50-338,.50-339. May 31,
1977. The subject matter was the inadequacy of the staf f review of safety issues relating to those f acilities .
- 2. Portland General Electric Co. (Trojan
. Nuclear Plant ) , Docket No. 50-344 (Control Building Proceedings ) , Dec . 11, 1978. The subject matter was the safety hazards of the nuclear plant and the inadequacy of NRC staff re view .
Respectfully submitted
~~
By. .
Ellyn R. Weiss SHELDON, HARMON & WEISS 1725 I S treet, N.W.
Suite 506 Washington, D.C. 20006 I hereby affirm that the above answers are true and accurate to the best of my knowledge and belief.
f) .<
-lA',o .;>p .'
n
- 2. -
7 ,4: / -
Robert D. Pollard Signed and sworn to-before me this -
day of ,, 7 6, 1980.
7-
. , ', c. ,1 /. .
NOTARY PUBLIC Mg Comm!s:ica 24m D :='m W P .
DATED: March-17, 1980
regaest should be n.at) withis the nort few renths that licensees con- cper-ttnq Procebres ider s$4ittoral (tatus amanitoring af virtaus ergineered safety fxtures nd their apporting servlees. no W Staff should begin attastes on the.
Svantages armi disadvantages of mach monitoring cri about the anne time Safety aspects of individual rea: tors darirg norm! operation and smsder
- als.
Aesponses from licensees should be er;ected in about one year, accident conditions are reviewed in detail by the W Staff and discussed a which time the W Staff should be in a position to review and evalte- with the A3S. Acceptable limits for norul o;= rations are form 11 sed by to them. Technical WRC staff. Specifications, santtted by the licensee arms a; proved by the Operating procafures for severe transients have received less so Cor sittee recognizas that some of the recornended actions in this detailed review by the W Staff. It a;' pears that such procesares would r; ort have already been taAan by the W Staff. benefit freen review by an interdisciplinary taan sAlch ine1LxSes personnel exprt both in operations and in systen behavior. Also, for the 1:nger f erely, term, there r.ay be merit in considerits; the developnent of more stedar$=
!*ed foru ts for mach procedures, v'N paltability of Electrie Pcur Supplies Max w. Carton Durtry the past several years there have been several operettig experi-csairman ences involving a loss of AC power to inqmrtant engineered safeguards.
Se AGS belteves it ir;mrtant that a coeprehensive reezminaticri be made by the AC arms the reactor licensees of the adequacy of design, testirug, arat minter.ince of offsite and cnsite AC and IC powr supplies. In par =
ticalar, f ailure endes ard effects analyses should be cade, if not al-rea$y prformm1, more systematic testirn of pwer systaus r:11 ability, in-notabla Joseph M. Bendrie cluaing atterral or an:malous system transients, should be considered, arut tirmian improved gaality assurance and status ronitorirq of powr arpply sysum should be rought.
- 5. Nuctsir Regulatofy Cirusission shlistan, DC 20555
^^*IYSI8 "I T^8I'"'8 bjects INTERIM FEPOT 10. 3 ON 2REE MIII iSIAND MC,iAR STATIC 38 WT2 rh Wt ead hsee aM Wr d a aswh Wt be asked to make a cetailed evaluation of his current capab!!!ty to Mth-er Dr* ReMriu stand station blackout (loss of offsite and onsite AC power) includirig additional complicating factors that alght be reasonably considered. Se tirn its 229th meetirq. My 10-12, 1979, the Advisory Ccrssittee cn evaluation should inc1 Lase examinaticn of natural circulatim capability, the continuiry availability of com;nnents needed for long-tern cooling, actor Safeguards continued its review of the recent accident at Dree and the potential for improvement in capability to survive extersled sta-le Islard Nuclear Station thit 2 (MI-2), ine1Lding (Eplicatlons drawn tion blackout.
a the occurrence of this accident.
nal r2comert$ations to make at this time.ne ccrarJttee has several am$di- T5e Acts also recocnends that each 11eensee and construct!an pruit holder should exmine a wide rarge of anomalous transients ard degram!as accident teter Presmre Vessel Level Indication 8 ' " #*
i l
preventing such conditions should be evaluated, as should researtti to
- Coneittee belteves,that it would to prufeat to-considat.exptottfously provide a better basis for such evaluations. 2e Ccunmittee expects it e proviticn oTEstrwentation thA wilIprovide an mambiguoes Irmilca- would be a;propriate to have mach studies done generically first, for n at the level of fluid in the reactor vessel. We arggest that licens- classes of reactor designs and system types.
e af all pressurized water reactore be requested to m2 sit design pro-als aruS ochetules for accorplishirq this action. 21s would assure h m Plann!ne timely availability of reviewd designs if the Staff ongoir:3 stusies uld indicate that early tr@!amentatton is required. he Ccrusittee An effort should be undertaken to plan and define the role NRC will play levcs tat as e einimum, the level indleation should range fzcza the in emergenetes ard eat their contributten and interactices will be with tm af the hot leg piping to the reactor vessel flange area. the 11ennsee and other energency plan prticipants ineltding other rf ter training and Nal!fication sant agencies, treustry representatives, armi national lateratories. govern-S=ch Plarnirq should cunsiders NRC Staff should emanine operator qualifications, trainirq. and 11~ ' .
assurarca that formal doementation of plans, procedures strq to detmsine mat charges are needed.
Consideration should be and organizaticn are in place for action in an energency, '
en to' educations! tackgrourts, to trainirq rethods, and to content of trctnirq progran. Attention should also te given to testing methods, . designaticn of a technical advisory team with names arut 2 epcific concern for the ability of the tasting rethods to predict alternates for the anticipated needs of an energency retor crpability. Examination of Ileensirn procedures should deter- situation,
- ether they are resronsive to new information that is developed it plant or orarator performance. Effort arculd also be made to .
comtletion of an inventory of equ!pwnt and materials traine
- ether remalts of emaninations can be correlated with oper- telch may be needed for unusual conditions inclia!! rug ability. Requalificatica trainirq and testirq abould be almilarly its description, location, availabt!!ty and the organ!-
sined ta insure that they take acco6mt of informatim that is devs1- sation ubicti cmtrols its release.
t by sporttica in the plant, ard to deterstne that relevant irtforne-i tbout sther plants is saia available to operators, ard is ma$e part We Ccranittee recorsse mis that each 11eerwee be asked to review and h3 traintrq ard repa1*:1 cation program. As part of this armi of revise within atout three amtlas r sore extensive atta$k1, continutry attention sust be given to the .
nt af infirmation thich an operator can assimilate arms use in rarmal his bases for obtainirn offsite advice and assistance la in emergency situations and to the best method of presentirq tae ire ernergerctes, frcus within ard outside the cmepary, atton to the oprator.
he use stir training should recotee ard consideration.
careful limitatters of simulators for .
current baus for rotifytrg and providing information to authorities ct' site in case of emergency.
2atten af ticense, twent pecorts his review and evaluation shtmld te in tarms of accidents havirq a i
2se af the pstentially valuable information contained in t.ieensee broad rarge of consegaenees. de results of this review should be l re;orted to the NRC.
l
- R ;a rts it.IAs), the Corsnittee reecreerufs that the NRC Staff estab-
, formal procedures for the use of this inferir.ation in the trainirg Decontamination aid Remvery j
[
! rpervisory arts maintenance staffs ass in the !!censtrq arms requali- <
- lon cf o;irattry fersonnel at correccial ruelear guwer plants. De !
I motion in Ins may also to useful in anticipatirq safety problema. me Corsnittee wishes to call attention to the 1eportance of a grogran de- 1 l
is prstent tire scee wt111ttes routinely request that they be pro- signed to learn directly abut the tehavior, failure modes, survivabil-t 1 copies af all tins queable to pJants of the type they operate ity, arsi other aspecs of com;onent and sptem tehavior at MI-2 as part l I
e speettic systens and e,rponents In a given clats of plants similar of the long-term recovery process. 21s progres should also exasine the ett plant. Certain reietor verdors have made similar requests and lessons learned at ?'I-2 to de ermine !! design cha ges are racessary to he IJAs to review ard evaluate the perforrarre of their plants. In facilitate the decontmination arms recovery of major nuclear ; owr plant systems.
! len, the AC operator lleensing staff has indicated that they use i in reviewirq c$erating esterlence at etrrercial fac!!!tles. j CC O "
argi rss%ber of (irs that attritnste the cause to personnel error }
R^ h P
terms ta indicate that a formalised pregram of Int review would be 1 in the training, lleensing a'tS requalificaticn of rurlear power personnel. We extent to which such a program could to used to
_ o oJ e J t J[ ~
ha!
j 1 pts safety problems should also be considered. 3 ;
a r,rv r ,yt,w proco$uree We ricomen5 that the Cmaission npest eam powr nector licensa and construction permit relder to perform design sta$les of a systam 21 cts e *.M1-2 accident has lagosed large row pressures on the evallability of adds the option of filtned ventim or Wm d contalmnt in W event of a serims ace ent. eys am a ou be TMe of wims@
n;ower risources within the Mc staff. If progress is to be erpedited ing a steam and Wgen envirceuwnt and of removing and utainig for thi new Westicns etch have arisen and m eatsting tmresolved safety as long a an as necessan radbaetive wtWates and W great 2 tser, the ACRS telieves r. hat riew mechanisms should be axqht and imple- of the iodine M accidents Mwig degW situations @ to and W nted. For those s.sfety concerns where such a machianism is a;propriata cluding cure melt. Such stta11es could be dane generically for several e Cawlttee recurands that the Comission should regaest licensees to reactor-contalment types, arus should evaluate the practicality, pros rfors sultat;1e atta$les on a timely tasis, inchm$1rq an evaluation of am5 cams, the casu, aa$ the totantial for risk redrtion. A perm M e pros and cms, aruf proposals for possible ispleventation of safety atat tw1w eonths for a report to the W by Heensees and constsch proverwnt . We AC staff se,muld concurrently establish its own cape- permit holders a; pears to represent a possible schedule.
11ty ta evalt.ata such sttsfies by arrars;1rq for support by its consult-ts atu! contractors. In this fashion, the Ccznittee anticipatas that e inforumtim cm thich jdpents will be based can te develo;md sas:h r3 careditiously, arms an earlier rem >1ution of many safety emcorns Ronorable .7osepti M. Her:frie y be achieved.
Chairman pability ef the NHC Staff' U. S. Mrlear Regulatory Cormission e Crsinittee reconsnenis that the capability of the 144C Staff to deal th basic arn$ ergineertrq protiens in what ray be termed broadly as
Subject:
REPCRI CN QUMTITAT1W SAFETY COA 1.1 actor at:1 fuel cycle cheatstry be aupanted erpeditiously. 21s ould include establisteent of expertise within the Isic, with assis* Dear Dr. Hendries tes tirarged frcus cxnsaltants armi contractors, in much inportant chntetl creas as the behevlor of IWR and IMR coolants arst 'other me- The Mvisory Coretittee on Reactor Safeguards recomernis that core-rials urder radiation cnnoittonas generation, hand 11 rag. and disposal sideration to given by the Nuclear Regulatory Cerunission to the tsololytic or other hydrogen at nuclear facilitiess ;=rformarce of establishrnent of quantitative safety goals for overall safety of rious chealcal alditives in contalrunent spraysa processirq and disposal nuclear powr reactors. %is could be helpful, for enaraple, in
<rtticpas for low aist high level radioactive wastess chemical operations develop 1 rug criteria for NRC actions concerning operatirq plants.
. other parts of the rauclear fuel cycler at:1 in the charmical treatment %e AmS reccqnizas the difficulties and mcertaintles in tre erstions involved in recovery, decontamination, or decornissionirrg of quantifleation of risk arut understarnis that in marry situations elser facilities. %e Canittee wishes to em;tsastas the 1sgortance of engireerirq ha$ pent will be the cnly or the primary basis for a ovidirq this errertise in toth the research aruS licanstrq management decision. Nevertheless, the Acts believes that the existence of ants af tha pac quantitative safety goals arn$ criteria can provide important yard-sticks for mach judpent.
ngle rsilare criterton The AmS believes that such NRC goals and criteria should be pro-
- e NRC should begin a sttsfy to determine if use of t!as sirqle (silure posed for corr ent, not only by the public but by the Congress.
Iterie establishes an a;propriate level of reliability for reactor Ult 1zately the Corgress should be asked to express its views on iftty systeps. Cteratirq ertettence sqgests that smaltiple failures the saltability of such goals arms criteria in relation to other si corron aise failures are encomtered with safficient frequency that relevant aspects of our technological society, such as large dass, eey need wore specific cxrtsideraticn. 21s study shaald be accompanied and manufacturissg, storage, or disposal facilities for hazardous a concurrtnt consideration of how the licenstrq pra=== can be sodified eg ,ge,y,,
e tas ecoount of a rww set of criteria as appropriaca.
The AmS believes that it is tine to place the discussion of risk, iftty wearch nuclear arms rennuclear, on as cpantitative a bests as feasible.
-e AmS believes that, as a result of the M1-2 accident, various safety gg ,, gy' esetrca cross wili warrent initiation or such greater evaphasis, as ap-opritte. %e Cornittee angests that consideraticn be given to an aug- ~
mtetton af the teC safety research taalget for rf 83.
Max W. Carbon Lee., the Ccrualttee believea that a larger part of the safety research g "
cgram should te oriented toward exploratory researcin as contrastad to
' infirmatery research, with see degree of freeds from issnedtate Itcens-g regattements. he AmS plans to I.sve a Subcormittee meetirq cn this abject with representatives of the NRC Office of Ntailcar Regulatory rsezrch in the rear future, w Carmittee is continu1rq to review these matters armi will report fur-Der ca mi!!tional rococanendatlans are developed.
33ttional csrvets by Messrs. E. Laus, D. Neller, D. Chrent, and J. Rey re prrsented belcn..
S neerely, -- - s
\1 n a om , ,
n.. N. C.rton h irman d L}a M~ w~
filtions! Corrents tw messrs. S. Iswis, D. Mneller, D. Okrent, and J. Ray Te potantial for a redaction in risk to the gublic in the case of a ser-ms reactor accident by the implerwntation of a seans for centrolled, 11tered wenttrq cf a ccntalrrwnt vitch could retain particulates and ie tulk of the icx!!ne has teen toesgnized for more than a decade. he
, neegt was recursonied for stta!y eore recently in tN Acerican Physical l
>ctsty Report on lightwater reactor sa fety arn$ in tie Ford roundattore-itra Report, *Nelear tower - 1swes ars! Ciotees.' It is a high pri-rIty teen in the *ac plan m2nitted to Congress for Research to In: prove se Safsty of LightWater Welear Pewer Plants (??JRIG0438). Se sttafy I
- rforred for the State of California cn trs!ergrotn$ sltirq concita$ad l
sat filtsted, vented cmntairrwnt ns a favored optirm to explore in can-l *etten with gmssible reans to mitigate the consequerres of serious re-l : tor sceldents. Fwever, little [vcgress has been sale on the develop-mt of raf f!clently detailed d.rsign informtion on teilch to evaluate the
( !!!cc:7 i tru! other factore relevant to a decision on gossible implementa-l Lon of c.ach cxrumquence snellorattrq systems.
I ie M1-2 a:cident mqgests that the probability of a serious accident in l
11 cts o filtered vented contattunent could to useful is larger than many
$3 artticipated.
4
.,r. . .
PERFORMANCE APPRAISAL AND RECORD OF INTERVIEN FOR NON-SUPERVISORY , .... , .
PROFESSIO:!AL TECHNICld. EMPLOYEE ,
~,
PROFILE .
N!01E: Robert D. Pollard GRADE / STEP: 14/2 PO';1 TION : Project Manager 1 / TIM'E IN GRADE: 19 =onths
- AI' PRAISE 9 h DA~E :
PEVIEWER & DATE:
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B as TIME IN STEP:
TIME IN PREVIOUS GRADE:
7 months 12 months s0 7.w R. E. "DcYpung g //-/5-h
~
PREVIOUS APPRAISER & DATE: D. B. Vassallo AEC/NRC Sr.RVICE (YEARS) : 6-l/4 11'/11/74 PROFESSIONAL EXPERIENCE (YEARS): 6 IIN! LONG SUPERVISED BY APPRAISER: 13 months .
DATE OF BIRTI:: 2/13/40 ar:
, EDUCATION (DECREE & YEAR): ,
P, ~'
B.S. (Elec. Eng.) 1969 o
Y DISCUSSION TO?ICS BACKCROUND & ENPERIENCE SUFDIARY - Bob received a B.S. in Electrical Engineering f rom Syracuse University in 1969. After joining the AEC in 1969, he studied ciectrical and nuclear engineering at the Graduate School of the University ,
of New Mexico (1970-1971) in conjunction with the AEC Intern Program. .;,;
Bob rerved for six years with the U. S. Navy as an electronic technician. "
lie served as an instructor, reactor operator, and was in charge of the ,,,
reactor control division aboard a nuclear-powered submarine. ....
Af ter joining the AEC in July 1969, Bob participated primarily in technical review groups in the review of instrumentation, control, and electrical .-
systems of nuclear power plants. For a brief period, he was a member of et the Standards group and participated in developing standards and safety '.
guides. He also served as a member of IEET, Cocmittecs. Bob transferred to RL as a project manager in September J.974.
- KNOWLEDGE OF JOB - Although Bob has excellent expertise in the instrumentation, ,
control, and electrical systems of nuclear power plants, he has also developed very good overall knowledge of nuclear power plant design. Since transferring ,
to L , he has shown the capability to rapidly expand his knowledge and under-standing of the diverse technical review areas with which a project manager .
must be familiar. Although he may_ require a littlu more exposure in certain review areas (e.g., auxiliary systems and site related matters), Bob is .
- . . . .a . . . ; . . . m . . . _ ... . .- . ,, .-. . .- . .
4
technically very perceptive. He has enough confidence to challenge reviewers f' on questionable technical matters and to pursue resolution of those in 7 cantroversy. ;
In a very short time, Bob has developed an excellent understanding of the technical, management, and administrative aspects of project management.
He manages to keep himscif informed of current developments in technical, policy, legal, and general licensing matters.
ECACCIENT CAPABILITY - Because.of.his pas.t experience in TR, Bob had a ,
good understanding of the LPM's role. .['
He has made a very rapid transition in assuming a project management philosophy. In the year that he has been I in RL, he has demonstrated an excellent capability to manage radiological safety reviews. Bob is an exceptionally thorough project manager who performs his tasks with a very critical view and in a very organized manner.
He is very knowledgeable of, and quick to grasp and implement administrative procedures. He is very effective in maintaining full cognizance of all aspects of his projects. He works very effectively with TR, OELD, and applicant representatives and gets along very well with the branch secretaries :,
and licensing assistant. '
,(
i Bob has shown excellent capability to effectively organize and manage ,
several concurrent projects. His principal assignment has been the OL review of the McGuire plant. However, he was also assigned as the LPM for the completion of the Comanche Peak and Catawba CP reviews. The latter required a considerable amount of LPM interaction with OELD because of the applicant's request for an exemption from meeting the ECCS criter,ia. Bob ,
showed great adeptness' at underst'anding and handling the unique technical -
legal aspects for bringing the Catawba project to an end; i.e. , issuance ,h
[.
of a CP. Because of the recent loss of an LPM from LWR l-1, Bob was .
g also assigned the task of completing the OL review of Indian Point 3, another project uith a long history of ' complexities. In handling all of ;'
these projects, Bob has shown a great deal of resourcefulness in moving *1 these projects forward concurrently without diminishing his efforts in any ...
one of them. .
PROFESSIONALISM - Bob is an extremely conscientious and dependable project '
manager. He conducts himself with a degree of maturity and professionalism <
well beyond his age. In his associations with applicant representatives, ;
he is very f air, but firm, and can'take a strong stance when the occasion .
warrants it. Bob does extremely well at planning and scheduling his
~
workload. He is consistently able to complets assignments on schedule s without needing reminders.
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W TT lh D 1 D' r !r L{Mw s Bob does not require much supervision. On the contrary, he seems to have ,
a unique instinct of knowing the t'ype of licensing action that a situation requires and then begins to take the appropriate action rithout waiting for direction frc the branch chief. In this regard, Eab has an out-standing knculedge of the Regulations and uorks very effectively uith lawyers (e.g. , has prepared soma quite involved technical - legal documents in conjunction with the Catawba and Indian Point- 3 projects) . He is very persistent in trying to get stalled actions moving. Bob does an excellent job of keeping his branch chief apprised of major review =ctters.
JUDCMENT - Bob is a careful thinker and uses good logic in making judgments. --
He has a very good understanding of the licensing progran and uses good .
judgment consistent with regulatory objectives.
CO?t-RTNI CATIO"S Oral - Bob has very good oral communicatica skills. He speaks clearly, with thought, and is very casily understood. He handles meetings extremely well. When he was a member of TR, he had considerabic neu experience and was very effective in presentations before the ACRS and was also exposed to public hearings. FC .
"*~
Written - Bob writes entremely well. The documents he prepares are concise and clear. As mentioned above, he has a decided instinct for knowing the type of action required and can translate this in writing without any apparent difficulty. His uritten work requires very little editing.
PERSONAL CHARACTEP.ISTICS - Basically, Bob is a very serious minded but --
persecable cmployec. He does not make rash decisions, but rather uses C ".E a more deliberative approach. Bob manages to maintain a rather even ,
composure no matter how difficult a situation may get.
Dob is an extremely conscientious, responsible, and dependable employee.
Occasionally he appears to become somewhat perplcned in rationalizing the ~
iralementation of licensing policy. In my opinion, this is because Bob has an exceptional understanding of the Cc= mission's rules and regulations and takcu his role of regulator very seriously. However, this has not af fected his performance as a project manager.
AREAS NCEDING IMPROVEMENT - Since transferring from TR, Bob is becoming exposed to a number of revicu areas with which he did not previously have a great deal of familiarity. These are principally in the areas of site safety, effluent treatment, and some portions of auxiliary systems. He ,
has mde great strides in understanding what the major revic objectives are for these areas. Uith the continued exposure he is now obtaining .
in managing his projects, 1 do not foresee any problem in Eob becoming q complete 1" conversaat in all review subjecta.
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PROMOTION POTE'!TIAL - Bob has shown excellent project canagement capability. -
IIe is well organized, is able to keep his projects under control, and to "
acet schedule milestones. On the basis of his previous e::perience in TR .
and with further c):perience in project management, Bob has an e:<cellent potential for attaining higher levels.
SU 0!ARY - Although Bob has been in RL for about one year, he has deccustratad cxcellent skills in managing safety reviews wi.thout requiring a great deal of supervision. Through his versatility, he has perfor=ed extremely well in handling di'icrse assigncents in a highly professional manner; e.g., ' -
taking on the management of cocple':i'c'ases'such as Catawba and Indian ,5 Point 3 in the final stages of licensing effort. ."
A REVIEWER'S CO C4E:iTS .
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Ai E}!PLOYEE'S CO}ME!!TS e
Acknowledce=cnt I have read the above performance appraisal.
'.'i f
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M Signaturc h [ /t? - I -
Date:
l$$7d&u.//tv /'}/?23 Corments by E=o lo'rce .
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
)
In the Matter of )
)
METROPOLITAN EDISON )
COMPANY, et al.,
~~ -~
) Docket No. 50-289
) (Restart)
(Three Mile Island )
Nuclear Station, Unit )
No. 1) )
)
CERTIFICATE OF SERVICE I hereby certify that a copy of the " Union of Concerned Scientists' Answers to Licensees First Set Of Interrogatories" were mailed first class postage pre-paid this 17th day of March, 1980 to the following parties:
Secretary of the Commission ATTN: Chief, Docketing and Service Section U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Ivan W. Smith, Esquire Atomic Safety & Licensing Board Panel U.S. Nuclear Regulatory Commission g Washington, D.C. 20555 e C Dr. Walter H. Jordan & g~~ $\
881 W. Outer Drive -\
Oak Ridge, Tennessee 37830 7)
L t!.}? \ ,~ '. -c > ~- a Dr. Linda W. Little ggedet,k][,f '.
5000 Hermitage Drive Ed.dM. ' ' ~ ;?
Raleiegh, North Carolina 27612 D'd48 N c, , . - -
l'
, George F. Trowbridge, Esquire Shaw, Pittman, Potts & Trowbridge 1800 "M" Street, N.W.
Washington, D.C. 20006 James Tourtellotte, Esquire Office of the Executive Legal Director .
U.S. Nuclear Regulatory Commission Washington, D.C. 20555 silyn x. weiss